Sample records for subcritical reactor systems

  1. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    Keywords. Accelerator driven systems; nuclear waste transmutation; computer codes; reactor physics; reactor noise; kinetics; burnup; transport theory; Monte Carlo; thorium utilization; neutron multiplication; sub-criticality; sub-critical facilities.

  2. Numerical simulations of flow field in the target region of accelerator-driven subcritical reactor system

    CERN Document Server

    Chen Hai Yan


    Numerical simulations of flow field were performed by using the PHOENICS 3.2 code for the proposed spallation target of accelerator-driven subcritical reactor system (ADS). The fluid motion in the target is axisymmetric and is treated as a 2-D steady-state problem. A body-fitted coordinate system (BFC) is then chosen and a two-dimensional mesh of the flow channel is generated. Results are presented for the ADS target under both upward and downward flow, and for the target with diffuser plate installed below the window under downward flow

  3. 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)


    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

  4. Neutronic design study of accelerator driven system (ADS) for Jordan subcritical reactor as a neutron source for nuclear research. (United States)

    Xoubi, Ned


    In this paper, a preliminary neutronic design study of an accelerator driven subcritical system for Jordan Subcritical Assembly (JSA) is presented. The conceptual design of coupling the JSA core with proton accelerator and spallation target is investigated, and its feasibility as a neutron source for nuclear research, and possibly for target irradiation and isotope production evaluated. 3D MCNPX model of the JSA reactor, the accelerator beam, and the Pb target was developed, based on actual reactor parameters. MCNPX calculations were carried out to estimate the absolute radial and axial neutron flux in the reactor, and to calculate the multiplication factor K eff and heat generated in the reactor. Numerical results showed an enormous increase in the neutron flux, by seven orders of magnitude, compared to the current JSA core design using Pu-Be source. In this research the results obtained are discussed and compared with those of the JSA, and do confirm the feasibility of utilizing the JSA as a viable nuclear research facility with adequate neutron flux. Copyright © 2017 Elsevier Ltd. All rights reserved.

  5. Neutrino Physics with Accelerator Driven Subcritical Reactors (United States)

    Ciuffoli, Emilio


    Accelerator Driven Subcritical System (ADS) reactors are being developed around the world, to produce energy and, at the same time, to provide an efficient way to dispose of and to recycle nuclear waste. Used nuclear fuel, by itself, cannot sustain a chain reaction; however in ADS reactors the additional neutrons which are required will be supplied by a high-intensity accelerator. This accelerator will produce, as a by-product, a large quantity of {\\bar{ν }}μ via muon Decay At Rest (µDAR). Using liquid scintillators, it will be possible to to measure the CP-violating phase δCP and to look for experimental signs of the presence of sterile neutrinos in the appearance channel, testing the LSND and MiniBooNE anomalies. Even in the first stage of the project, when the beam energy will be lower, it will be possible to produce {\\bar{ν }}e via Isotope Decay At Rest (IsoDAR), which can be used to provide competitive bounds on sterile neutrinos in the disappearance channel. I will consider several experimental setups in which the antineutrinos are created using accelerators that will be constructed as part of the China-ADS program.

  6. 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)


    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)

  7. 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


    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)

  8. 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)


    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)

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

    Indian Academy of Sciences (India)

    made plutonium are the two key elements that are serving as nuclear fuel. Nuclear power presently constitutes about 17% of the total electric power generation in the world from about 430 operating reactors. Although this figure of global nuclear energy generation appears modest, the share of nuclear electricity in several ...

  10. The physics of accelerator driven sub-critical reactors

    Indian Academy of Sciences (India)

    utilization; neutron multiplication; sub-criticality; sub-critical facilities. PACS Nos 89.30.Gg; 28.41.-I; 28.50.-k. 1. Introduction. Accelerator driven systems (ADS) are attracting worldwide attention increasingly due to their superior safety characteristics and their potential for burning actinide and fission product-waste and energy ...

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

    Blaise, Patrick; Mellier, Frédéric; Fougeras, Philippe


    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.

  12. 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:; Huang, Kai; He, Mingtao; Li, Xunzhao


    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroshi


    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.

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

    Energy Technology Data Exchange (ETDEWEB)

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


    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

  15. 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)


    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)

  16. 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: [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)


    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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


    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.

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

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


    Ebrahimkhani, Marziye; Hassanzadeh, Mostafa; Feghhi, Sayed Amier Hossian; Masti, Darush


    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 ...

  20. 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)


    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)

  1. 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: [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)


    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

  2. First reactivity determination of a subcritical reactor using a single beam-trip and fission chambers operating in current mode

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Ordonez, M.; Villamarin, D.; Becares, V.; Gonzalez-Romero, E.M. [Nuclear Innovation Group, CIEMAT, Avda. Complutense, Madrid (Spain); Bergloef, C. [Reactor Physics Department, Royal Institute of Technology, Stockholm (Sweden); Bournos, V.; Fokov, Y.; Kiyavitskaya, H. [Joint Institute for Power and Nuclear Research, National Academy of Sciences, Minsk (Belarus)


    Transmutation of spent nuclear fuel in Accelerator-Driven Systems (ADS) is considered as a key technology for achieving sustainable nuclear energy. In the design of future ADS facilities, the reactivity monitoring system is of highest importance. An extensive experimental program devoted to reactivity monitoring of ADS has been carried out at the subcritical facility YALINA-Booster in the framework of IP-EUROTRANS. The main objective, besides the qualification of the reactivity monitoring techniques, has been to develop electronic chains that can be used in a full power ADS. For this purpose, YALINA-Booster couples a D-T neutron generator to a flexible zero-power subcritical assembly with a coupled fast-thermal neutron spectrum. The high intensity of the accelerator and the possibility to work in continuous or pulsed mode allowed the study of the current-to-flux relationship and beam-trip experiments. In addition, the experimental facility provided the opportunity to test electronic chains in current mode, which correspond to the most probable condition in a full power ADS. There exists a relationship between the reactivity of a subcritical core, the intensity of the accelerator and the neutron source intensity. Hence, by monitoring these three quantities it should be possible to determine the origin of any reactivity or power change within the subcritical assembly. We have developed the necessary acquisition system to monitor the conditions of these three variables in the millisecond scale. The current-to-flux technique provides relative changes in the behavior of the core, however, in order to determine absolute values of the reactivity, we have taken profit of short imposed beam interruptions in the millisecond scale, thus providing the possibility for applying the Source-Jerk method within few milliseconds. It is the first time that the reactivity of an ADS is determined in a single beam-trip using fission chambers operating in current mode. The experiments

  3. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications; Implementacao e qualificacao de metodologia de calculos neutronicos em reatores subcriticos acionados por fonte externa de neutrons e aplicacoes

    Energy Technology Data Exchange (ETDEWEB)

    Carluccio, Thiago


    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{sub eff} and k{sub 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)

  4. Microwave ion source for accelerator driven sub-critical system

    CERN Document Server

    Cui Bao Qun; Jiang Wei; LiLiQiang; WangRongWen


    A microwave ion source is under developing for a demonstration prototype of a accelerator driven sub-critical system at CIAE (China Institute of Atomic Energy), 100 mA hydrogen beam has been extracted from the source through a 7.3 mm aperture in diameter, proton ratio is more than 85%, reliability has been tested for 100 h without any failures

  5. Monte Carlo Alpha Iteration Algorithm for a Subcritical System Analysis

    Directory of Open Access Journals (Sweden)

    Hyung Jin Shim


    Full Text Available The α-k iteration method which searches the fundamental mode alpha-eigenvalue via iterative updates of the fission source distribution has been successfully used for the Monte Carlo (MC alpha-static calculations of supercritical systems. However, the α-k iteration method for the deep subcritical system analysis suffers from a gigantic number of neutron generations or a huge neutron weight, which leads to an abnormal termination of the MC calculations. In order to stably estimate the prompt neutron decay constant (α of prompt subcritical systems regardless of subcriticality, we propose a new MC alpha-static calculation method named as the α iteration algorithm. The new method is derived by directly applying the power method for the α-mode eigenvalue equation and its calculation stability is achieved by controlling the number of time source neutrons which are generated in proportion to α divided by neutron speed in MC neutron transport simulations. The effectiveness of the α iteration algorithm is demonstrated for two-group homogeneous problems with varying the subcriticality by comparisons with analytic solutions. The applicability of the proposed method is evaluated for an experimental benchmark of the thorium-loaded accelerator-driven system.

  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: [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)


    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. Hydrolysis of polycarbonate in sub-critical water in fused silica capillary reactor with in situ Raman spectroscopy (United States)

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


    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. ?? 2009 The Royal Society of Chemistry.

  8. The TRADE experiment: shielding calculations for the building hosting the subcritical system. (United States)

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


    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.

  9. On feasibility of optimizing the neutronic parameters of a laser system pumped by a pulsed reactor


    A.V. Gulevich; O.F. Kukharchuk; A.I. Brezhnev; A.A. Suvorov


    The paper examines the calculated feasibility of improving the energy characteristics of power pulses in a system consisting of a reactor and a subcritical block. A BARS-type fast neutron reactor is used as a self-quenching pulsed reactor. The subcritical block is a cylindrical structure comprising laser-active elements, moderator components and two reflectors (internal and external). The internal reflector material is zirconium hydride, and the external reflector material is beryllium. Th...

  10. Implementation and qualification of neutronic calculation methodology in subcritical reactors driven by external neutron sources and applications


    Thiago Carluccio


    O trabalho teve como objetivo a investigação de Metodologias de Cálculo dos Reatores Subcríticos acionados por fonte externa de nêutrons, tais como, \\"Accelerator Driven Subcritical Reactor\\" (ADSR) e \\"Fusion Driven Subcritical Reator\\" (FDSR) , que são reatores nucleares subcríticos com uma fonte externa de nêutrons. Tais nêutrons são produzidos, no caso do ADSR, através da interação de partículas aceleradas (prótons, deutério) com um alvo (Pb, Bi, etc) ou através das reações de fusão, no c...

  11. Thorium as a Fuel for Accelerator Driven Subcritical Electronuclear Systems

    CERN Document Server

    Barashenkov, V S; Singh, V


    Neutron yield and energy production in a very large, practically infinite, uranium and thorium target-blocks irradiated by protons with energies in the range 0.1-2 GeV are studied by Monte Carlo method. Though the comparison of uranium and thorium targets shows that the neutron yield in the latter is 30-40 % less and the energy gain is approximatelly two times smaller, accelerator Driven subcritical Systems (ADS) with thorium fuel are very perspective at the bombarding energies higher than several hundreds MeV. An admixture of fissile elements U^{233}, U^{235}, Pu^{239} in the set-up gives larger neutron multiplication which in turn shows better energy amplification. It is argued that due to the practically complete burning of the fuel in such set-up there is no need of technology of conversion of the exhaust fuel.

  12. The Chain-Length Distribution in Subcritical Systems

    Energy Technology Data Exchange (ETDEWEB)

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


    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.

  13. 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


    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.


    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.


    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.


    Treshow, M.


    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  16. 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: [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080-A (Venezuela, Bolivarian Republic of)


    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)

  17. Superconducting RF Linacs Driving Subcritical Reactors for Profitable Disposition of Surplus Weapons-grade Plutonium (United States)

    Cummings, Mary Anne; Johnson, Rolland

    Acceptable capital and operating costs of high-power proton accelerators suitable for profitable commercial electric-power and process-heat applications have been demonstrated. However, studies have pointed out that even a few hundred trips of an accelerator lasting a few seconds would lead to unacceptable thermal stresses as each trip causes fission to be turned off in solid fuel structures found in conventional reactors. The newest designs based on the GEM*STAR concept take such trips in stride by using molten-salt fuel, where fuel pin fatigue is not an issue. Other aspects of the GEM*STAR concept which address all historical reactor failures include an internal spallation neutron target and high temperature molten salt fuel with continuous purging of volatile radioactive fission products such that the reactor contains less than a critical mass and almost a million times fewer volatile radioactive fission products than conventional reactors. GEM*STAR is a reactor that without redesign will burn spent nuclear fuel, natural uranium, thorium, or surplus weapons material. It will operate without the need for a critical core, fuel enrichment, or reprocessing making it an excellent candidate for export. As a first application, the design for a pilot plant is described for the profitable disposition of surplus weapons-grade plutonium by using process heat to produce green diesel fuel for the Department of Defense (DOD) from natural gas and renewable carbon.

  18. A pragmatic approach towards designing a second shutdown system for Tehran research reactor


    Boustani Ehsan; Khakshournia Samad; Khalafi Hossein


    One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall reliability of the reactor shutdown system. The proposed second shutdown system is a diverse, independent shutdown system compared to the existing rod based one that intends to achieve and maintain sub-criticality condition with an enough shutdown margin in man...

  19. Equilibria of ternary system acetic acid-water-CO2 under subcritical conditions

    DEFF Research Database (Denmark)

    Gutierrez, Jose M. Jimenez; Mussatto, Solange I.; Tsou, Joana

    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......) of the ternary system HAc—H2O—CO2 at different subcritical conditions. A proposed computer model could be validated with experimental data, leading to a certain degree of adjustment due to specific factors, such as the binary interaction parameter kij, used in the model based on the Peng-Robinson EoS coupled...... 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...

  20. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.


    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...... 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...

  1. Subcritical nuclear assembly

    Energy Technology Data Exchange (ETDEWEB)

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


    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)

  2. A fusion-driven subcritical system concept based on viable technologies (United States)

    Wu, Y.; Jiang, J.; Wang, M.; Jin, M.; FDS Team


    A fusion-driven hybrid subcritical system (FDS) concept has been designed and proposed as spent fuel burner based on viable technologies. The plasma fusion driver can be designed based on relatively easily achieved plasma parameters extrapolated from the successful operation of existing fusion experimental devices such as the EAST tokamak in China and other tokamaks in the world, and the subcritical fission blanket can be designed based on the well-developed technologies of fission power plants. The simulation calculations and performance analyses of plasma physics, neutronics, thermal-hydraulics, thermomechanics and safety have shown that the proposed concept can meet the requirements of tritium self-sufficiency and sufficient energy gain as well as effective burning of nuclear waste from fission power plants and efficient breeding of nuclear fuel to feed fission power plants.

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

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim


    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.

  4. 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)


    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.

  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


    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 on design of superconducting proton linac for accelerator driven subcritical nuclear power system

    CERN Document Server

    Yu Qi; Xu Tao Guang


    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 (SCL) is a good selection of ADS accelerator because of its high efficiency and low beam loss rate. It is constitute by a series of the superconducting accelerating cavities. The cavity geometry is determined by means of the electromagnetic field computation. The SCL main parameters are determined by the particle dynamics computation

  7. Nuclear reactor sealing system (United States)

    McEdwards, James A.


    A liquid metal-cooled nuclear reactor sealing system. 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.

  8. Accelerator driven sub-critical core (United States)

    McIntyre, Peter M; Sattarov, Akhdiyor


    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.

  9. Attrition reactor system (United States)

    Scott, Charles D.; Davison, Brian H.


    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  10. 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


    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

  11. 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: [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080A (Venezuela, Bolivarian Republic of)


    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)

  12. Modeling of the CTEx subcritical unit using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Avelino [Divisao de Defesa Quimica, Biologica e Nuclear. Centro Tecnologico do Exercito - CTEx, Guaratiba, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: [Programa de Engenharia Nuclear. Universidade Federal do Rio de Janeiro - UFRJ Centro de Tecnologia, Rio de Janeiro, RJ (Brazil); Rebello, Wilson F. [Secao de Engenharia Nuclear - SE/7 Instituto Militar de Engenharia - IME Rio de Janeiro, RJ (Brazil); Cunha, Victor L. Lassance [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)


    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{sub eff}, a higher concentration of U235 can be proposed, provided it safely remains below one. (author)

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


    Peng Hu; Gao-Wei Zhang; Long-Xiang Chen; Ming-Hou Liu


    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 analys...

  14. 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


    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)

  15. Subcritical multiplication determination studies

    Energy Technology Data Exchange (ETDEWEB)

    Estes, G.P.; Goulding, C.A.


    A series of measurements and improvements to computational techniques are in progress at Los Alamos National Laboratory that are aimed at better understanding the determination of the reactivity of subcritical systems from measurements of the apparent multiplication of the system. Such studies are being performed in order to improve the special nuclear material (SNM) assays of unknown systems such as those encountered in SNM safeguards, arms-control verification, imports of foreign-generated SNM, etc. Improved techniques and understanding are needed since measured multiplication is not always an invariant characteristic of a subcritical system, especially if one has a system with no significant intrinsic internal neutron source that is illuminated nonuniformly with an external source (i.e., a non-normal mode system).

  16. CFD Analysis and Design of Detailed Target Configurations for an Accelerator-Driven Subcritical System

    Energy Technology Data Exchange (ETDEWEB)

    Kraus, Adam; Merzari, Elia; Sofu, Tanju; Zhong, Zhaopeng; Gohar, Yousry


    High-fidelity analysis has been utilized in the design of beam target options for an accelerator driven subcritical system. Designs featuring stacks of plates with square cross section have been investigated for both tungsten and uranium target materials. The presented work includes the first thermal-hydraulic simulations of the full, detailed target geometry. The innovative target cooling manifold design features many regions with complex flow features, including 90 bends and merging jets, which necessitate three-dimensional fluid simulations. These were performed using the commercial computational fluid dynamics code STAR-CCM+. Conjugate heat transfer was modeled between the plates, cladding, manifold structure, and fluid. Steady-state simulations were performed but lacked good residual convergence. Unsteady simulations were then performed, which converged well and demonstrated that flow instability existed in the lower portion of the manifold. It was established that the flow instability had little effect on the peak plate temperatures, which were well below the melting point. The estimated plate surface temperatures and target region pressure were shown to provide sufficient margin to subcooled boiling for standard operating conditions. This demonstrated the safety of both potential target configurations during normal operation.

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

    Directory of Open Access Journals (Sweden)

    Zhihui Li


    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.

  18. The study of isochoric subcritical water using power series: A potential of energy generation with ISCW reactor (United States)

    Sangian, Hanny F.; Tunena, Mercyas; Pani, Sutaryono


    The present study was aimed to analyze the behaviors of PVT and Z of ISCW (isochoric subcritical water) condition using mathematical series. The data showed that the pressure extremely increased from 15 bars until 80 bars taking only a few seconds that was probably to generate energy. The study was initiated by formulating power series in term of P and T with parameters, a0, a1, a2, a3, a4, a5, a6, and a7 whereas they were solved by performing the fitting method. By employing that technique, parameters were obtained as follows: a0= 7.63x-6, a1=0.23K/bar, a2=0.0035K2/bar2, a3=0.0068 K3/bar3, a4=8.27x10-7K4/bar4, a5=1.33x10-7K5/bar5, a6=2.18x10-8K6/bar6, and a7=3.64x10-9K7/bar7. Compressibility factor increased as pressure and temperature improved. In an isochoric condition, there was extremity that was located at a temperature above 450K in which compressibility factor abruptly increased with the tangent of the line was infinity. All parameters attaching on terms of a mathematical model proposed were assumed constant during P and T increase. At the request of the authors of the paper and with the agreement of the proceedings editor, an updated version of this article was published on 4 October 2017. The original version supplied to AIP Publishing included an incorrect spelling in the name of the first author. This has been corrected in the updated and re-published version.

  19. 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: [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of)


    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)

  20. Conceptual design of minor actinides burner with an accelerator-driven subcritical system.

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Y.; Gohar, Y. (Nuclear Engineering Division)


    In the environmental impact study of the Yucca Mountain nuclear waste repository, the limit of spent nuclear fuel (SNF) for disposal is assessed at 70,000 metric tons of heavy metal (MTHM), among which 63,000 MTHM are the projected SNF discharge from U.S. commercial nuclear power plants though 2011. Within the 70,000 MTHM of SNF in storage, approximately 115 tons would be minor actinides (MAs) and 585 tons would be plutonium. This study describes the conceptual design of an accelerator-driven subcritical (ADS) system intended to utilize (burn) the 115 tons of MAs. The ADS system consists of a subcritical fission blanket where the MAs fuel will be burned, a spallation neutron source to drive the fission blanket, and a radiation shield to reduce the radiation dose to an acceptable level. The spallation neutrons are generated from the interaction of a 1 GeV proton beam with a lead-bismuth eutectic (LBE) or liquid lead target. In this concept, the fission blanket consists of a liquid mobile fuel and the fuel carrier can be LBE, liquid lead, or molten salt. The actinide fuel materials are dissolved, mixed, or suspended in the liquid fuel carrier. Therefore, fresh fuel can be fed into the fission blanket to adjust its reactivity and to control system power during operation. Monte Carlo analyses were performed to determine the overall parameters of an ADS system utilizing LBE as an example. Steady-state Monte Carlo simulations were studied for three fission blanket configurations that are similar except that the loaded amount of actinide fuel in the LBE is either 5, 7, or 10% of the total volume of the blanket, respectively. The neutron multiplication factor values of the three configurations are all approximately 0.98 and the MA initial inventories are each approximately 10 tons. Monte Carlo burnup simulations using the MCB5 code were performed to analyze the performance of the three conceptual ADS systems. Preliminary burnup analysis shows that all three conceptual ADS

  1. Alpha Eigenvalue Estimation from Dynamic Monte Carlo Calculation for Subcritical Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shaukat, Nadeem; Shim, Hyung Jin; Jang, Sang Hoon [Seoul National University, Seoul (Korea, Republic of)


    The dynamic Monte Carlo (DMC) method has been used in the TART code for the α eigenvalue calculations. A unique method has been equipped to measure the α in time-stepwise Monte Carlo simulations. For off-critical systems, the neutron population is allowed to change exponentially over a period of time. The neutron population is uniformly combed to return it to the neutron population started with at the beginning of time boundary. In this study, the conventional dynamic Monte Carlo method has been implemented in the McCARD. There is an exponential change of neutron population at the end of each time boundary for off-critical systems. In order to control this exponential change at the end of each time boundary, a conventional time cut-off controlling population strategy is included in the DMC module implemented in the McCARD. the conventional combing method to control the neutron population for off-critical systems is implemented. Instead of considering the cycles, the simulation is divided in time intervals. At the end of each time interval, neutron population control is applied on the banked neutrons. Randomly selected neutrons are discarded, until the size of neutron population matches the initial neutron histories at the beginning of time simulation. The prompt neutron decay constant α is estimated from DMC algorithm for subcritical systems. The effectiveness of the results is examined for two-group infinite homogeneous problems with varying the k-value. From the comparisons with the analytical solutions, it is observed that the results are quite comparable with each other for each k-value.

  2. Solvent refined coal reactor quench system (United States)

    Thorogood, R.M.


    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. 1 fig.

  3. Rapid starting methanol reactor system (United States)

    Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.


    The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

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

    Directory of Open Access Journals (Sweden)

    Peng Hu


    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.

  5. Automatically scramming nuclear reactor system (United States)

    Ougouag, Abderrafi M.; Schultz, Richard R.; Terry, William K.


    An automatically scramming nuclear reactor system. One embodiment comprises a core having a coolant inlet end and a coolant outlet end. A cooling system operatively associated with the core provides coolant to the coolant inlet end and removes heated coolant from the coolant outlet end, thus maintaining a pressure differential therebetween during a normal operating condition of the nuclear reactor system. A guide tube is positioned within the core with a first end of the guide tube in fluid communication with the coolant inlet end of the core, and a second end of the guide tube in fluid communication with the coolant outlet end of the core. A control element is positioned within the guide tube and is movable therein between upper and lower positions, and automatically falls under the action of gravity to the lower position when the pressure differential drops below a safe pressure differential.

  6. Liquid metal cooled nuclear reactor plant system (United States)

    Hunsbedt, Anstein; Boardman, Charles E.


    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  7. Subcritical and supercritical water oxidation of organic, wet wastes for carbon cycling in regenerative life support systems (United States)

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


    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

  8. Shutdown system for a nuclear reactor (United States)

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.


    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  9. Optimization of supercritical phase and combined supercritical/subcritical conversion of lignocellulose for hexose production by using a flow reaction system. (United States)

    Zhao, Yan; Lu, Wen-Jing; Wu, Hua-Yong; Liu, Jin-Wen; Wang, Hong-Tao


    A flow reaction system was utilized to investigate lignocellulose conversion using combined supercritical/subcritical conditions for hexose production. Initially, investigation of cellulose hydrolysis in supercritical water and optimization of reaction parameters were done. Oligosaccharide yields reached over 30% at cellulose concentrations of 3-5 gL(-1) and reaction times of 6-10s at 375 °C, and 2.5-4 gL(-1) and 8-10s at 380 °C. Temperatures above 380 °C were not appropriate for the supercritical phase in the combined process. Subsequently, conversion of lignocellulosic materials under combined supercritical/subcritical conditions was studied. Around 30% hexose was produced from corn stalks under the optimal parameters for supercritical (380 °C, 23-24 MPa, 9-10s) and subcritical (240 °C, 8-9 MPa, 45-50s) phases. Flow systems utilizing the combined supercritical/subcritical technology present a promising method for lignocellulosic conversion. The results of this study provide an important guide for the operational optimization and practical application of the proposed system. Copyright © 2012 Elsevier Ltd. All rights reserved.

  10. On the significance of the energy correlations of spallation neutrons on the neutron fluctuations in accelerator-driven subcritical systems

    CERN Document Server

    Pázsit, I; Fhager, V


    Studies of neutron fluctuations in spallation-driven subcritical systems require the use of energy-dependent master equations. In particular, calculation of the second moment of the neutron distribution requires knowledge on the energy correlations (two-point distributions) of the source particles. It is shown here that such correlations will exist even if the energies of all neutrons, generated in any single spallation event, are independent, provided that the energy distribution of the neutrons for separate spallation events is dependent on the number of neutrons generated. A simple model of number dependence of the energy spectrum is constructed, and the arising energy correlations are calculated. The error in calculating the second moment of the neutron distribution, arising when assuming zero correlations (i.e. using only one-particle energy spectra), is estimated in a simple model of neutron slowing down.

  11. Design, Development and Installation of Jordan Subcritical Assembly


    Ned Xoubi


    Following its announcement in 2007 to pursue a nuclear power program and in the absence of any nuclear facility essential for the education, training, and research, Jordan decided to build a subcritical reactor as its first nuclear facility. Jordan Subcritical Assembly (JSA) is uranium fueled light water moderated and reflected subcritical reactor driven by a plutonium-beryllium source, and the core consists of 313 LEU fuel rods, loaded into a water-filled vessel in a square lattice of 19.11 ...

  12. Tandem Mirror Reactor Systems Code (Version I)

    Energy Technology Data Exchange (ETDEWEB)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.


    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.

  13. Nuclear electric propulsion reactor control systems status (United States)

    Ferg, D. A.


    The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

  14. Fission control system for nuclear reactor (United States)

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  15. 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)


    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.

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


    Daniel Maraver; Sylvain Quoilin; Javier Royo


    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 deman...

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

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Johanna


    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.

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

    Indian Academy of Sciences (India)

    ... as well as for nuclear energy generation utilizing thorium as fuel. In India, there is an interest in the programmes of development of high-energy and high-current accelerators due to the potential of ADS in utilizing the vast resources of thorium in the country for nuclear power generation. The accelerator related activities ...

  19. A pragmatic approach towards designing a second shutdown system for Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Boustani Ehsan


    Full Text Available One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall reliability of the reactor shutdown system. The proposed second shutdown system is a diverse, independent shutdown system compared to the existing rod based one that intends to achieve and maintain sub-criticality condition with an enough shutdown margin in many of abnormal situations. It is designed as much as practical based on neutron absorber solution injection into the existing core while the changes and interferences with the existing core structure are kept to a minimum. Core neutronic calculations were performed using MCNPX 2.6.0 and MTR_PC package for the current operational core equipped with the second shutdown system, and one experiment was conducted in the Tehran Research Reactor to test the neutronic calculations. A good agreement was seen between theoretical results and experimental ones. In addition, capability of the second shutdown system in the case of occurrence of design basis accident in the Tehran Research Reactor is demonstrated using PARET program.

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


    Jae-Yong Lim; Cheol Ho Pyeon; Takahiro Yagi; Tsuyoshi Misawa


    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...

  1. Cooling system for a nuclear reactor (United States)

    Amtmann, Hans H.


    A cooling system for a gas-cooled nuclear reactor is disclosed which includes at least one primary cooling loop adapted to pass coolant gas from the reactor core and an associated steam generator through a duct system having a main circulator therein, and at least one auxiliary cooling loop having communication with the reactor core and adapted to selectively pass coolant gas through an auxiliary heat exchanger and circulator. The main and auxiliary circulators are installed in a common vertical cavity in the reactor vessel, and a common return duct communicates with the reactor core and intersects the common cavity at a junction at which is located a flow diverter valve operative to effect coolant flow through either the primary or auxiliary cooling loops.

  2. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth


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

    Energy Technology Data Exchange (ETDEWEB)



    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)

  4. 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


    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.

  5. Scanning tunneling microscope assembly, reactor, and system (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A


    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.

  6. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar


    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.


    Miller, G.


    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W


    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.

  9. Monte Carlo simulation of a perturbed subcritical core

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, Mustafa K.; Park, Chang Je [KAERI, Daejeon (Korea, Republic of)


    Jordan Subcritical Assembly (JSA) is designed for the purpose of education, training, and experiment research. Jordan subcritical assembly is considered Jordan's First Nuclear Facility Moving Jordan into the nuclear age. It is a teaching and training experimental facility that is designed to stay in a subcriticality A subcritical assembly is a multiplying system of nuclear fuel and moderator whose effective multiplication factor is less than unity. An extraneous source of neutron is required for the operation in order to compensate for the difference between the production rate of fission neutrons in the fuel and the rate of loss caused by absorption and leakage.

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

    Directory of Open Access Journals (Sweden)

    Fang Yan


    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.


    Moore, W.T.


    This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

  12. Transients in reactors for power systems compensation (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

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

    DEFF Research Database (Denmark)

    Ommen, Torben; Elmegaard, Brian


    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...

  14. RF gauging efforts with liquid hydrogen and liquid oxygen as applicable to subcritical space vehicle systems (United States)

    Thompson, H. E.; Ott, W.; Stanley, N.


    The RF gauging concept is based on the interaction between a fluid dielectric medium in an enclosed metallic cavity and electromagnetic fields set up within that cavity. In RF gauging systems, the fundamental measurement relies on the interpretation of changes in the resonant RF frequencies of an enclosed tank as the mass of the propellant contained in the tank is changed. In addition to a discussion of the basic principles of operation of these systems, the study presents a description of the current breadboard implementation with typical test arrangements, along with supporting test data. The experimental testing of the RF gauging technique for liquid cryogen mass gauging indicates that this technique is a feasible approach to liquid oxygen and liquid hydrogen gauging under all attitude or reduced gravity environments.

  15. Laser fusion power reactor system (LFPRS)

    Energy Technology Data Exchange (ETDEWEB)

    Kovacik, W. P.


    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. (MOW)

  16. Subcritical neutron production using multiple accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, W.Y.; Jones, J.L. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Harmon, J.F. [Idaho State Univ., Pocatello, ID (United States)


    A subcritical neutron production technique using multiple accelerators is being developed to provide a selective alternative (for small volumes) to nuclear reactor neutron production. The concept combines the capabilities of multiple commercially-available linear accelerators and a compact subcritical assembly design to generate reactor-like thermal neutron fluxes (i.e., 10{sup 13}-10{sup 14} n/cm{sup 2}/s) in small irradiation volumes of up to 500 cm{sup 3}. In addition, fast and epithermal neutron fluxes will also be available. The neutron source utilizes radially-oriented, pulsed, electron accelerators. The subcritical neutron production assembly is in the form of a compact right-cylinder (approximately 20-cm dia.). This assembly uses an outer ring of graphite (i.e., reflector) with re-entrant holes to enable penetration of the electron beam to the internal structure which comprises of uranium as an electron convertor/neutron multiplier followed by H{sub 2}O beryllium, H{sub 2}O aluminum, and D{sub 2}O in succession toward the center. The inner-most region filled with D{sub 2}O is the central irradiation volume. The material configuration and overall design is to maximize thermal neutron fluxes in the central irradiation volume based on photoneutron/photofission and neutron multiplication processes as well as neutron transport. This assembly will be designed not to reach a nuclear critical state under any normal and/or accidental condition.

  17. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

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


    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...... and malfunctions of anaerobic digesters occasionally experienced, leading to sub-optimal methane production and wastewater treatment. Using a variety of molecular techniques, we are able to determine which microorganisms are active, where they are active, and when they are active, but we still need to determine...

  18. 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


    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.

  19. Rodded shutdown system for a nuclear reactor (United States)

    Golden, Martin P.; Govi, Aldo R.


    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. 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.

  20. Hybrid Molten Salt Reactor (HMSR) System Study

    Energy Technology Data Exchange (ETDEWEB)

    Woolley, Robert D [PPPL; Miller, Laurence F [PPPL


    Can the hybrid system combination of (1) a critical fission Molten Salt Reactor (MSR) having a thermal spectrum and a high Conversion Ratio (CR) with (2) an external source of high energy neutrons provide an attractive solution to the world's expanding demand for energy? The present study indicates the answer is an emphatic yes.

  1. Control system studies for thermionic reactors (United States)

    Hermsen, R. J.; Gronroos, H. G.


    In core thermionic reactor concepts are of interest for space missions that require electrical power in the range of a few tens of kilowatts up to several megawatts. The physical principle involved--thermionic direct conversion of heat to electricity at net efficiencies up to 15 percent--offers potential advantages when compared to other nuclear powerplant concepts. However, the integration of the thermionic diode electrode structure with high-temperature nuclear fuel materials presents new design problems and new reactor physical constraints. Among the topics that must be investigated are those associated with the control system. The results of analytical and simulation studies of thermionic reactor control performed at the Jet Propulsion Laboratory are discussed.

  2. Reactor power system deployment and startup (United States)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.


    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.

  3. 10 CFR 50.46a - Acceptance criteria for reactor coolant system venting systems. (United States)


    ... 10 Energy 1 2010-01-01 2010-01-01 false Acceptance criteria for reactor coolant system venting... criteria for reactor coolant system venting systems. Each nuclear power reactor must be provided with high point vents for the reactor coolant system, for the reactor vessel head, and for other systems required...

  4. 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


    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.

  5. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing (United States)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.


    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

  6. Quantum Subcritical Bubbles (United States)

    Uesugi, T.; Morikawa, M.; Shiromizu, T.


    We quantize subcritical bubbles which are formed in the weakly first order phase transition. We find that the typical size of the thermal fluctuation reduces in quantum-statistical physics. We estimate the typical size and the amplitude of thermal fluctuations near the critical temperature in the electroweak phase transition using a quantum statistical average. Furthermore, based on our study, we discuss implications for the dynamics of phase transitions.

  7. Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system (United States)

    Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun


    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.

  8. Basic concept for an accelerator-driven subcritical system to be used as a long-pulse neutron source for Condensed Matter research

    Energy Technology Data Exchange (ETDEWEB)

    Vivanco, R., E-mail: [ESS-BILBAO, Parque Tecnológico Bizkaia, Laida Bidea, Edificio 207 B Planta Baja, 48160 Derio (Spain); Instituto de Fusión Nuclear - UPM, ETS Ingenieros Industriales, C/ José Gutiérrez Abascal, 2, 28006 Madrid Spain (Spain); Ghiglino, A.; Vicente, J.P. de; Sordo, F.; Terrón, S.; Magán, M. [ESS-BILBAO, Parque Tecnológico Bizkaia, Laida Bidea, Edificio 207 B Planta Baja, 48160 Derio (Spain); Instituto de Fusión Nuclear - UPM, ETS Ingenieros Industriales, C/ José Gutiérrez Abascal, 2, 28006 Madrid Spain (Spain); Perlado, J.M. [Instituto de Fusión Nuclear - UPM, ETS Ingenieros Industriales, C/ José Gutiérrez Abascal, 2, 28006 Madrid Spain (Spain); Bermejo, F.J. [Instituto de Estructura de la Materia, IEM-CSIC, Consejo Superior de Investigaciones Científicas, Serrano 123, 28006 Madrid (Spain)


    A model for an accelerator-driven subcritical system to be operated as a source of cold neutrons for Condensed Matter research is developed at the conceptual level. Its baseline layout relies upon proven accelerator, spalattion target and fuel array technologies, and consists in a proton accelerator able to deliver some 67.5 mA of proton beam with kinetic energy 0.6 GeV, a pulse length of 2.86 ms, and repetition rate of 14 Hz. The particle beam hits a target of conventional design that is surrounded by a multiplicative core made of fissile/fertile material, composed by a subcritical array of fuel bars made of aluminium Cermet cooled by light water poisoned with boric acid. Relatively low enriched uranium is chosen as fissile material. An optimisation of several parameters is carried out, using as components of the objective function several characteristics pertaining the cold neutron pulse. The results show that the optimal device will deliver up to 80% of the cold neutron flux expected for some of the ongoing projects using a significantly lower proton beam power than that managed in such projects. The total power developed within the core rises up to 22.8 MW, and the criticality range shifts to a final k{sub eff} value of around 0.9 after the 50 days cycle.

  9. Integral reactor system and method for fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Neil Edward; Brown, Michael S.; Cheekatamaria, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F.


    A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert higher 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.

  10. Microchannel Reactor System for Catalytic Hydrogenation

    Energy Technology Data Exchange (ETDEWEB)

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie


    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.

  11. Reactor Design for Bioelectrochemical Systems

    KAUST Repository

    Mohanakrishna, G.


    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.

  12. Laser fusion hybrid reactor systems study

    Energy Technology Data Exchange (ETDEWEB)


    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 (less lasers).

  13. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system (United States)

    Harto, Andang Widi


    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  14. Power conditioning for space nuclear reactor systems (United States)

    Berman, Baruch


    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.

  15. Staged membrane oxidation reactor system (United States)

    Repasky, John Michael; Carolan, Michael Francis; Stein, VanEric Edward; Chen, Christopher Ming-Poh


    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. Concept of the power-reactor-pumped laser for technology applications (United States)

    Gulevich, Andrey V.; Dyachenko, Peter P.; Kononov, Victor N.; Kukharchuk, Oleg F.; Zrodnikov, Anatoly V.


    Conception of a high-power pulsed reactor-pumped laser system (RPLS) based on new physical principles (direct nuclear-to- optical energy conversion) for the technology and space application is discussed. The development of an energy model of RPLS consisting of the ignition two-core fast-burst reactor reactor module and a thermal subcritical laser module filled with an Ar-Xe laser active medium is reported. Some of the experimental results are also presented.

  17. 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)


    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.

  18. Nuclear reactor pressure vessel support system (United States)

    Sepelak, George R.


    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  19. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,


    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.

  20. Weld monitor and failure detector for nuclear reactor system (United States)

    Sutton, Jr., Harry G.


    Critical but inaccessible welds in a nuclear reactor system are monitored throughout the life of the reactor by providing small aperture means projecting completely through the reactor vessel wall and also through the weld or welds to be monitored. The aperture means is normally sealed from the atmosphere within the reactor. Any incipient failure or cracking of the weld will cause the environment contained within the reactor to pass into the aperture means and thence to the outer surface of the reactor vessel where its presence is readily detected.

  1. Nuclear power history calculation for subcritical systems using Euler-MacLaurin formula; Calculo do historico de potencia nuclear para sistemas subcriticos utilizando a formula de Euler-MacLaurin

    Energy Technology Data Exchange (ETDEWEB)

    Henrice Junior, Edson; Goncalves, Alessandro da Cruz, E-mail:, E-mail: [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear; Palma, Daniel Arthur Pinheiro, E-mail: [Comissao Nacional de Energia Nucleara (CNEN), Rio de Janeiro, RJ (Brazil)


    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)


    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski


    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.

  3. 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)


    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.

  4. 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


    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.

  5. Radiolytic production of chemical fuels in fusion reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Fish, J D


    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.

  6. Dynamics of Subcritical Bubbles in First Order Phase Transition (United States)

    Shiromizu, T.; Morikawa, M.; Yokoyama, J.


    We derivate the Langevin and the Fokker-Planck equations for the radius of O(3)-symmetric subcritical bubbles as a phenomenological model to treat thermal fluctuation. The effect of thermal noise on subcritical bubbles is examined. We find that the fluctuation-dissipation relation holds and that in the high temperature phase the system settles down rapidly to the thermal equilibrium state even if it was in a nonequilibrium state initially. We then estimate the typical size of subcritical bubbles as well as the amplitude of fluctuations on that scale. We also discuss their implication to the electroweak phase transition.

  7. MCNP multiplication analysis of subcritical HEU experiments

    Energy Technology Data Exchange (ETDEWEB)

    Estes, G.P. [Los Alamos National Lab., NM (United States); Brockhoff, R.C. [Kansas State Univ., Manhattan, KS (United States)


    A series of measurements and improvements to computational techniques was described in Ref. 1 that were aimed at better understanding the determination of the reactivity of subcritical systems from measurements of the multiplying characteristics of the system. This methodology has been applied to a number of the bare highly enriched uranium (HEU) measurements (simulating 0.5- to 21.5-kg balls with nesting shells) of Ref. 2, demonstrating that the experimental multiplication results can be reproduced computationally with good accuracy. This capability promises to improve special nuclear material (SNM) assays of unknown systems such as those encountered in SNM safeguards, arms-control verification, imports of foreign-generated SNM, smuggling of SNM, etc. Improved techniques and understanding are needed since traditionally measured or calculated multiplications are not always an invariant characteristic of a subcritical system, especially if one has an SNM system with no significant intrinsic internal neutron source that is illuminated nonuniformly with an external source (i.e., a nonnormal mode system). The measurement techniques used in Refs. 1 and 2 to determine multiplication are based on the Feynman variance-to-mean method, which has been previously documented in Refs. 3 and 4 and applied successfully to normal mode systems such as plutonium and uranium spheres. These techniques have been applied to nonnormal mode problems with less success, and both Refs. 1 and 2 as well as the current paper are attempts to better understand the subcritical multiplication of such systems.

  8. Integrated systems analysis of the PIUS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others


    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  9. Nuclear reactor system study for NASA/JPL (United States)

    Palmer, R. G.; Lundberg, L. B.; Keddy, E. S.; Koenig, D. R.


    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.

  10. Evaluation of reactivity monitoring techniques at the Yalina - Booster sub-critical facility

    Energy Technology Data Exchange (ETDEWEB)

    Becares Palacios, V.


    The management of long-lived radioactive wastes produced by nuclear reactors constitutes one of the main challenges of nuclear technology nowadays. A possible option for its management consists in the transmutation of long lived nuclides into shorter lived ones. Accelerator Driven Subcritical Systems (ADS) are one of the technologies in development to achieve this goal. An ADS consists in a subcritical nuclear reactor maintained in a steady state by an external neutron source driven by a particle accelerator. The interest of these systems lays on its capacity to be loaded with fuels having larger contents of minor actinides than conventional critical reactors, and in this way, increasing the transmutation rates of these elements, that are the main responsible of the long-term radiotoxicity of nuclear waste. One of the key points that have been identified for the operation of an industrial-scale ADS is the need of continuously monitoring the reactivity of the subcritical system during operation. For this reason, since the 1990s a number of experiments have been conducted in zero-power subcritical assemblies (MUSE, RACE, KUCA, Yalina, GUINEVERE/FREYA) in order to experimentally validate these techniques. In this context, the present thesis is concerned with the validation of reactivity monitoring techniques at the Yalina-Booster subcritical assembly. This assembly belongs to the Joint Institute for Power and Nuclear Research (JIPNR-Sosny) of the National Academy of Sciences of Belarus. Experiments concerning reactivity monitoring have been performed in this facility under the EUROTRANS project of the 6th EU Framework Program in year 2008 under the direction of CIEMAT. Two types of experiments have been carried out: experiments with a pulsed neutron source (PNS) and experiments with a continuous source with short interruptions (beam trips). For the case of the first ones, PNS experiments, two fundamental techniques exist to measure the reactivity, known as the prompt

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

    Energy Technology Data Exchange (ETDEWEB)

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


    A /sup 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 /sup 252/Cf totaling slightly over 100 mg. These sources are placed in a 93 percent /sup 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/sup +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.

  12. Accelerator-Reactor Coupling for Energy Production in Advanced Nuclear Fuel Cycles (United States)

    Heidet, Florent; Brown, Nicholas R.; Haj Tahar, Malek

    This article is a review of several accelerator-reactor interface issues and nuclear fuel cycle applications of accelerator-driven subcritical systems. The systems considered here have the primary goal of energy production, but that goal is accomplished via a specific application in various proposed nuclear fuel cycles, such as breed-and-burn of fertile material or burning of transuranic material. Several basic principles are reviewed, starting from the proton beam window including the target, blanket, reactor core, and up to the fuel cycle. We focus on issues of interest, such as the impact of the energy required to run the accelerator and associated systems on the potential electricity delivered to the grid. Accelerator-driven systems feature many of the constraints and issues associated with critical reactors, with the added challenges of subcritical operation and coupling to an accelerator. Reliable accelerator operation and avoidance of beam trips are critically important. One interesting challenge is measurement of blanket subcriticality level during operation. We also review the potential benefits of accelerator-driven systems in various nuclear fuel cycle applications. Ultimately, accelerator-driven subcritical systems with the goal of transmutation of transuranic material have lower 100,000-year radioactivity than a critical fast reactor with recycling of uranium and plutonium.

  13. Mechanical systems development of integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Chang, M. H.; Kim, J. I.; Choi, S.; Kim, K. S.; Kim, T. W.; Jeong, K. H.; Kim, J. H.; Kim, Y. W.; Lee, G. M.


    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose applications such as small capacity power generation, co-generation and sea water desalination. This in mind, survey has been made on the worldwide small and medium integral reactors under development. Reviewed are their technical characteristics, development status, design features, application plans, etc. For the mechanical design scope of work, the structural concept compatible with the characteristics and requirements of integral reactor has been established. Types of major components were evaluated and selected. Functional and structural concept, equipment layout and supporting concept within the reactor pressure vessel have also been established. Preliminary mechanical design requirements were developed considering the reactor lifetime, operation conditions, and the expected loading combinations. To embody the concurrent design approach, recent CAD technology and team engineering concept were evaluated. (author). 31 refs.,16 tabs., 35 figs.

  14. DNA-Based Enzyme Reactors and Systems

    Directory of Open Access Journals (Sweden)

    Veikko Linko


    Full Text Available During recent years, the possibility to create custom biocompatible nanoshapes using DNA as a building material has rapidly emerged. Further, these rationally designed DNA structures could be exploited in positioning pivotal molecules, such as enzymes, with nanometer-level precision. This feature could be used in the fabrication of artificial biochemical machinery that is able to mimic the complex reactions found in living cells. Currently, DNA-enzyme hybrids can be used to control (multi-enzyme cascade reactions and to regulate the enzyme functions and the reaction pathways. Moreover, sophisticated DNA structures can be utilized in encapsulating active enzymes and delivering the molecular cargo into cells. In this review, we focus on the latest enzyme systems based on novel DNA nanostructures: enzyme reactors, regulatory devices and carriers that can find uses in various biotechnological and nanomedical applications.

  15. Systems aspects of a space nuclear reactor power system (United States)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.


    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  16. High Performance Photocatalytic Oxidation Reactor System Project (United States)

    National Aeronautics and Space Administration — Pioneer Astronautics proposes a technology program for the development of an innovative photocatalytic oxidation reactor for the removal and mineralization of...

  17. Design, Development and Installation of Jordan Subcritical Assembly

    Directory of Open Access Journals (Sweden)

    Ned Xoubi


    Full Text Available Following its announcement in 2007 to pursue a nuclear power program and in the absence of any nuclear facility essential for the education, training, and research, Jordan decided to build a subcritical reactor as its first nuclear facility. Jordan Subcritical Assembly (JSA is uranium fueled light water moderated and reflected subcritical reactor driven by a plutonium-beryllium source, and the core consists of 313 LEU fuel rods, loaded into a water-filled vessel in a square lattice of 19.11 mm pitch. The fuel rods are based on PWR fuel structural pattern type, made of uranium oxide (UO2 with 3.4 wt% 235U enrichment in zirconium alloy (Zr-4 cladding. Design, optimization, and verification were performed using MCNP5 nuclear code; the computed effective multiplication factor is 0.95923. The JSA is designed to fulfill the training needs of students and is equipped to perform all of the fundamental experiments required for a typical nuclear engineering university program. This paper presents the design, development, modeling, core analysis, and utilization of Jordan’s first nuclear facility and why this simplified low cost reactor presents an attractive choice to fulfill the preliminary experimental needs of nuclear engineering education in developing countries.

  18. REACTOR - a Concept for establishing a System-of-Systems (United States)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim


    REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

  19. An analysis system for in-reactor behavior, FANTASI

    Energy Technology Data Exchange (ETDEWEB)

    Uto, Nariaki; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sugaya, Toshio; Sakai, Kimiaki [Japan Nucler Cycle Developmnet Inst., Tokai, Ibaraki (Japan)


    The Japan Nuclear Fuel Cycle Development Institute developed FANTASI (A Computational System For Analyzing Coupled Neutronic, Thermal-Hydraulic And Structural Behaviors In A Fast Breeder Reactor Core) to simulate a conditions where nuclear reaction, thermal-hydraulic behavior of coolant and deformation of core construction progress under mutual relation in reactor of a fast breeder reactor by cooperation of engineers in the fields of physics, thermal-hydraulics, structure, and information system on reactor. Here was described on system construction of FANTASI after describing progress of this development. And then, after introducing a case study using this system, applicability to transient phenomena in nuclear reactor was described. At last, with summarizing results of this development, its future development was also mentioned. (G.K.)

  20. 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


    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

  1. Passive modular gas safety system for a reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abalin, S.S.; Isaev, I.F.; Kulakov, A.A.; Sivokon, V.P.; Udovenko, A.N.; Ionaitis, R.R.


    Reactor safety systems have developed gradually. Today in particular, auxiliary systems are being developed which are based on nontraditional operational concepts, by using gaseous neutron absorbers. The Scientific-Research and Design Institute of Power Technology (NIKIET) and the Institute of Nuclear Reactors, Kurchatov Institute Reactor Science Center (RNTs), have done preliminary development and experimental verification of separate elements of this system, in which helium is used as the absorber. This article presents a rapid passive safety system based on gaseous absorber, which is made as autonomous modules as the final stage of reactor safety. Its effectiveness is discussed by using an RBMK reactor as an example. As opposed to traditional active, systems, it does not require a functioning power supply and information signals from outside the reactors system, which makes it stable against unsanctioned actions by personnel, the influence of other systems, and also outside actions (sabotage and natural calamities which could destroy the the nuclear power plant structure). Because the gas safety system can operate instantaneously (0.1-0.3 sec), in principle, it can shut down the reactor even with fast-neutron runaway, where traditional safety systems are ineffective.

  2. SP-100 Program: space reactor system and subsystem investigations

    Energy Technology Data Exchange (ETDEWEB)

    Harty, R.B.


    For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs.

  3. A new VFA sensor technique for anaerobic reactor systems

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær


    to monitor VFA online in one of the most difficult media: animal slurry or manure. A novel in situ filtration technique has made it possible to perform microfiltration inside a reactor system. This filter enables sampling from closed reactor systems without large-scale pumping and filters. Furthermore, due...... to its small size it can be placed in lab-scale reactors without disturbing the process. Using this filtration technique together with commercially available membrane filters we have constructed a VFA sensor system that can perform automatic analysis of animal slurry at a frequency as high as every 15...... filtration technique are being presented is this article....

  4. Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems (United States)

    Was, Gary S.


    The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems.

  5. Reactor Lithium Heat Pipes for HP-STMCs Space Reactor Power System (United States)

    Tournier, Jean-Michel; El-Genk, Mohamed S.


    Design and performance analysis of the nuclear reactor's lithium heat pipes for a 110-kWe Heat Pipes-Segmented Thermoelectric Module Converters (HP-STMCs) Space Reactor Power system (SRPS) are presented. The evaporator length of the heat pipes is the same as the active core height (0.45 m) and the C-C finned condenser is of the same length as the STMC panels (1.5 m). The C-C finned condenser section is radiatively coupled to the collector shoes of the STMCs placed on both sides. The lengths of the adiabatic section, the values of the power throughput and the evaporator wall temperature depend on the radial location of the heat pipe in the reactor core and the number and dimensions of the potassium heat pipes in the heat rejection radiator. The reactor heat pipes have a total length that varies from 7.57 to 7.73 m, and a 0.2 mm thick Mo-14%Re wick with an average pore radius of 12 μm. The wick is separated from the Mo-14%Re wall by a 0.5 mm annulus filled with liquid lithium, to raise the prevailing capillary limit. The nominal evaporator (or reactor) temperature varies from 1513 to 1591 K and the thermal power of the reactor is 1.6 MW, which averages 12.7 kW for each of the 126 reactor heat pipes. The power throughput per heat pipe increase to a nominal 15.24 kW at the location of the peak power in the core and to 20.31 kW when an adjacent heat pipe fails. The prevailing capillary limit of the reactor heat pipes is 28.3 kW, providing a design margin >= 28%.

  6. Emergency heat removal system for a nuclear reactor (United States)

    Dunckel, Thomas L.


    A heat removal system for nuclear reactors serving as a supplement to an Emergency Core Cooling System (ECCS) during a Loss of Coolant Accident (LOCA) comprises a plurality of heat pipes having one end in heat transfer relationship with either the reactor pressure vessel, the core support grid structure or other in-core components and the opposite end located in heat transfer relationship with a heat exchanger having heat transfer fluid therein. The heat exchanger is located external to the pressure vessel whereby excessive core heat is transferred from the above reactor components and dissipated within the heat exchanger fluid.

  7. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)


    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  8. Autonomous Control of Space Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na


    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  9. Development study on subcriticality monitor. 1. Report under business contract with Japan Nuclear Fuel Cycle Development Institute

    CERN Document Server

    Yamada, S


    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...

  10. A Design of Alarm System in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants.

  11. Digital, remote control system for a 2-MW research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Battle, R.E.; Corbett, G.K.


    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs.

  12. Plasma heating systems planned for the Argonne experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bertoncini, P.; Brooks, J.; Fasolo, J.; Mills, F.; Moretti, A.; Norem, J.


    A scoping study and conceptual design of a tokamak experimental power reactor (TEPR) have been completed. The design objectives of the TEPR are to operate for ten years at or near electrical power breakeven conditions with a duty factor of greater than or equal to 50 percent and to demonstrate the feasibility of tokamak fusion power reactor techniques. These objectives can be met by a design which has a major radius of 6.25 m and a plasma radius of 2.1 m. Parameters for this reactor are listed, and a diagram is given. This paper will describe TEPR plasma heating systems. Neutral beam heating and rf heating are described.

  13. Off-Design Performances of Subcritical and Supercritical Organic Rankine Cycles in Geothermal Power Systems under an Optimal Control Strategy

    National Research Council Canada - National Science Library

    Tieyu Gao; Changwei Liu


    .... In this study, an off-design performance prediction model for geothermal ORC systems is developed according to special designs of critical components, and an optimal control strategy which regards...

  14. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire


    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  15. Deployment history and design considerations for space reactor power systems (United States)

    El-Genk, Mohamed S.


    The history of the deployment of nuclear reactors in Earth orbits is reviewed with emphases on lessons learned and the operation and safety experiences. The former Soviet Union's "BUK" power systems, with SiGe thermoelectric conversion and fast neutron energy spectrum reactors, powered a total of 31 Radar Ocean Reconnaissance Satellites (RORSATs) from 1970 to 1988 in 260 km orbit. Two of the former Soviet Union's TOPAZ reactors, with in-core thermionic conversion and epithermal neutron energy spectrum, powered two Cosmos missions launched in 1987 in ˜800 km orbit. The US' SNAP-10A system, with SiGe energy conversion and a thermal neutron energy spectrum reactor, was launched in 1965 in 1300 km orbit. The three reactor systems used liquid NaK-78 coolant, stainless steel structure and highly enriched uranium fuel (90-96 wt%) and operated at a reactor exit temperature of 833-973 K. The BUK reactors used U-Mo fuel rods, TOPAZ used UO 2 fuel rods and four ZrH moderator disks, and the SNAP-10A used moderated U-ZrH fuel rods. These low power space reactor systems were designed for short missions (˜0.5 kW e and ˜1 year for SNAP-10A, <3.0 kW e and <6 months for BUK, and ˜5.5 kW e and up to 1 year for TOPAZ). The deactivated BUK reactors at the end of mission, which varied in duration from a few hours to ˜4.5 months, were boosted into ˜800 km storage orbit with a decay life of more than 600 year. The ejection of the last 16 BUK reactor fuel cores caused significant contamination of Earth orbits with NaK droplets that varied in sizes from a few microns to 5 cm. Power systems to enhance or enable future interplanetary exploration, in-situ resources utilization on Mars and the Moon, and civilian missions in 1000-3000 km orbits would generate significantly more power of 10's to 100's kW e for 5-10 years, or even longer. A number of design options to enhance the operation reliability and safety of these high power space reactor power systems are presented and discussed.

  16. New reactor technology: safety improvements in nuclear power systems. (United States)

    Corradini, M L


    Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

  17. TREAT (Transient Reactor Test Facility) reactor control rod scram system simulations and testing

    Energy Technology Data Exchange (ETDEWEB)

    Solbrig, C.W.; Stevens, W.W.


    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs.

  18. Sub-critical long-term operation of industrial scale hollow-fibre membranes in a submerged anaerobic MBR (HF-SAnMBR) system


    Robles Martínez, Ángel; Ruano García, María Victoria; Ribes Bertomeu, José; Ferrer, J.


    The aim of this study was to evaluate the long-term performance of hollow-fibre (HF) membranes used to treat urban wastewater in a submerged anaerobic MBR when operating sub-critically. To this end, a demonstration plant with two industrial scale HF ultrafiltration membrane modules was operated under different conditions. The main factor affecting membrane performance was the concentration of mixed liquor total solids (MLTS). The reversible fouling rate remained low even when MLTS levels (abo...

  19. Modeling and simulation of CANDU reactor and its regulating system (United States)

    Javidnia, Hooman

    Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

  20. Design of virtual SCADA simulation system for pressurized water reactor (United States)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman


    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  1. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail:; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)


    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  2. Simplified safety and containment systems for the iris reactor

    Energy Technology Data Exchange (ETDEWEB)

    Conway, L.E. [Westinghouse Electric Co., Pittsburgh, PA (United States); Lombardi, C.; Ricotti, M.; Oriani, L. [Polytechnic of Milan, Dept. of Nuclear Engineering, Milan (Italy)


    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  3. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    Energy Technology Data Exchange (ETDEWEB)

    Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.


    Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

  4. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)


    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  5. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

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


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

  6. Small reactor power systems for manned planetary surface bases (United States)

    Bloomfield, Harvey S.


    A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

  7. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.


    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  8. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman


    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  9. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.


    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  10. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)


    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  11. An introduction to the engineering of fast nuclear reactors

    CERN Document Server

    Judd, Anthony M


    An invaluable resource for both graduate-level engineering students and practising nuclear engineers who want to expand their knowledge of fast nuclear reactors, the reactors of the future! This book is a concise yet comprehensive introduction to all aspects of fast reactor engineering. It covers topics including neutron physics; neutron flux spectra; flux distribution; Doppler and coolant temperature coefficients; the performance of ceramic and metal fuels under irradiation, structural changes, and fission-product migration; the effects of irradiation and corrosion on structural materials, irradiation swelling; heat transfer in the reactor core and its effect on core design; coolants including sodium and lead-bismuth alloy; coolant circuits; pumps; heat exchangers and steam generators; and plant control. The book includes new discussions on lead-alloy and gas coolants, metal fuel, the use of reactors to consume radioactive waste, and accelerator-driven subcritical systems.

  12. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri


    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  13. Subcritical calculation of the nuclear material warehouse;Calculo de subcriticidad del almacen del material nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, T.; Mazon R, R., E-mail: teodoro.garcia@inin.gob.m [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)


    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)

  14. System and method for temperature control in an oxygen transport membrane based reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.


    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  15. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)


    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  16. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)


    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

  17. Fault detection system for Argentine Research Reactor instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Polenta, H.P. (Argentine Navy, Comodoro Py 2055 Office 11-93, 1104 - Buenos Aires (Argentina)); Bernard, J.A. (Nuclear Reactor Laboratory, Massachusetts Institute of Technology, 138 Albany Street, Cambridge, Massachusetts 02139 (United States)); Ray, A. (205 Mechanical Engineering Department, Pennsylvania State University, University Park, Pennsylvania 16802 (United States))


    The design and implementation of a redundancy management scheme for the on-line detection and isolation of faulty sensors is presented. Such a device is potentially useful in reactor-powered spacecraft for enhancing the processing capabilities of the main computer. The fault detection device can be used as an integral part of intelligent instrumentation systems. The device has been built using an 8-bit microcontroller and commercially available electronic hardware. The software is completely portable. The operation of this device has been successfully demonstrated for real-time validation of sensor data on Argentina's RA-1 Research Reactor.

  18. Space-reactor electric systems: subsystem technology assessment

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, R.V.; Bost, D.; Determan, W.R.


    This report documents the subsystem technology assessment. For the purpose of this report, five subsystems were defined for a space reactor electric system, and the report is organized around these subsystems: reactor; shielding; primary heat transport; power conversion and processing; and heat rejection. The purpose of the assessment was to determine the current technology status and the technology potentials for different types of the five subsystems. The cost and schedule needed to develop these potentials were estimated, and sets of development-compatible subsystems were identified.

  19. Modification of the Core Cooling System of TRIGA 2000 Reactor (United States)

    Umar, Efrizon; Fiantini, Rosalina


    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  20. Development of essential system technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Y. Y.; Hwang, Y. D.; Cho, B. H. and others


    Basic design of SMART adopts the new advanced technologies which were not applied in the existing 1000MWe PWR. However, the R and D experience on these advanced essential technologies is lacking in domestic nuclear industry. Recently, a research on these advanced technologies has been performed as a part of the mid-and-long term nuclear R and D program, but the research was limited only for the small scale fundamental study. The research on these essential technologies such as helically coiled tube steam generator, self pressurizer, core cooling by natural circulation required for the development of integral reactor SMART have not been conducted in full scale. This project, therefore, was performed for the development of analysis models and methodologies, system analysis and thermal hydraulic experiments on the essential technologies to be applied to the 300MWe capacity of integral reactor SMART and the advanced passive reactor expected to be developed in near future with the emphasis on experimental investigation. (author)

  1. Coupled Subcritical Water and Solid Phase Extraction for In-Situ Chemical Analysis Project (United States)

    National Aeronautics and Space Administration — Leiden Measurement Technology (LMT) will design and develop a low volume analyte separation, concentration, and transfer system (ConTech), that couples a Subcritical...

  2. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.


    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  3. Study of Natural Convection Passive Cooling System for Nuclear Reactors (United States)

    Abdillah, Habibi; Saputra, Geby; Novitrian; Permana, Sidik


    Fukushima nuclear reactor accident occurred due to the reactor cooling pumps and followed by all emergencies cooling systems could not work. Therefore, the system which has a passive safety system that rely on natural laws such as natural convection passive cooling system. In natural convection, the cooling material can flow due to the different density of the material due to the temperature difference. To analyze such investigation, a simple apparatus was set up and explains the study of natural convection in a vertical closed-loop system. It was set up that, in the closed loop, there is a heater at the bottom which is representing heat source system from the reactor core and cooler at the top which is showing the cooling system performance in room temperature to make a temperature difference for convection process. The study aims to find some loop configurations and some natural convection performances that can produce an optimum flow of cooling process. The study was done and focused on experimental approach and simulation. The obtained results are showing and analyzing in temperature profile data and the speed of coolant flow at some point on the closed-loop system.

  4. Subcritical and supercritical water oxidation of CELSS model wastes (United States)

    Takahashi, Y.; Wydeven, T.; Koo, C.


    A mixture of ammonium hydroxide with acetic acid and a slurry of human feces, urine, and wipes were used as CELSS model wastes to be wet-oxidized at temperatures from 250 to 500 C, i.e. below and above the critical point of water (374 C and 218 kg/sq cm or 21.4 MPa). The effects of oxidation temperature ( 250-500 C) and residence time (0-120 mn) on carbon and nitrogen and on metal corrosion from the reactor material were studied. Almost all of the organic matter in the model wastes was oxidized in the temperature range from 400 to 500 C, above the critical conditions for water. In contrast, only a small portion of the organic matter was oxidized at subcritical conditions. A substantial amount of nitrogen remained in solution in the form of ammonia at temperatures ranging from 350 to 450 C suggesting that, around 400 C, organic carbon is completely oxidized and most of the nitrogen is retained in solution. The Hastelloy C-276 alloy reactor corroded during subcritical and supercritical water oxidation.

  5. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    Ahmmed Saadi Ibrehem


    Full Text Available A modified model for the neutralization process of Stirred Tank Reactors (CSTR reactor is presented in this study. The model accounts for the effect of strong acid [HCL] flowrate and strong base [NaOH] flowrate with the ionic concentrations of [Cl-] and [Na+] on the Ph of the system. In this work, the effect of important reactor parameters such as ionic concentrations and acid and base flowrates on the dynamic behavior of the CSTR is investigated and the behavior of mathematical model is compared with the reported models for the McAvoy model and Jutila model. Moreover, the results of the model are compared with the experimental data in terms of pH dynamic study. A good agreement is observed between our model prediction and the actual plant data. © 2011 BCREC UNDIP. All rights reserved(Received: 1st March 2011, Revised: 28th March 2011; Accepted: 7th April 2011[How to Cite: A.S. Ibrehem. (2011. Modified Mathematical Model For Neutralization System In Stirred Tank Reactor. Bulletin of Chemical Reaction Engineering & Catalysis, 6(1: 47-52. doi:10.9767/bcrec.6.1.825.47-52][How to Link / DOI: || or local: ] | View in 

  6. Saphyr: a code system from reactor design to reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)


    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.

  7. Expert system for online surveillance of nuclear reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.


    This report describes an expert system for online surveillance of nuclear reactor coolant pumps. This system provides a means for early detection of pump or sensor degradation. Degradation is determined through the use of a statistical analysis technique, sequential probability ratio test, applied to information from several sensors which are responsive to differing physical parameters. The results of sequential testing of the data provide the operator with an early warning of possible sensor or pump failure.

  8. Operation of staged membrane oxidation reactor systems (United States)

    Repasky, John Michael


    A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

  9. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors (United States)


    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  10. Proliferation Potential of Accelerator-Drive Systems: Feasibility Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Riendeau, C.D.; Moses, D.L.; Olson, A.P.


    Accelerator-driven systems for fissile materials production have been proposed and studied since the early 1950s. Recent advances in beam power levels for small accelerators have raised the possibility that such use could be feasible for a potential proliferator. The objective of this study is to review the state of technology development for accelerator-driven spallation neutron sources and subcritical reactors. Energy and power requirements were calculated for a proton accelerator-driven neutron spallation source and subcritical reactors to produce a significant amount of fissile material--plutonium.

  11. Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India (United States)

    Sinha, Amar; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P. S.; Bishnoi, Saroj


    The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated keff of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor ks and external neutron source efficiency φ∗ in great details. Experiments with D-T neutrons are also underway.

  12. Summary of space nuclear reactor power systems, 1983--1992

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.


    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  13. Monitoring system for a liquid-cooled nuclear fission reactor (United States)

    DeVolpi, Alexander


    A monitoring system for detecting changes in the liquid levels in various regions of a water-cooled nuclear power reactor, viz., in the downcomer, in the core, in the inlet and outlet plenums, at the head, and elsewhere; and also for detecting changes in the density of the liquid in these regions. A plurality of gamma radiation detectors are used, arranged vertically along the outside of the reactor vessel, and collimator means for each detector limits the gamma-radiation it receives as emitting from only isolated regions of the vessel. Excess neutrons produced by the fission reaction will be captured by the water coolant, by the steel reactor walls, or by the fuel or control structures in the vessel. Neutron capture by steel generates gamma radiation having an energy level of the order of 5-12 MeV, whereas neutron capture by water provides an energy level of approximately 2.2 MeV, and neutron capture by the fission fuel or its cladding provides an energy level of 1 MeV or less. The intensity of neutron capture thus changes significantly at any water-metal interface. Comparative analysis of adjacent gamma detectors senses changes from the normal condition with liquid coolant present to advise of changes in the presence and/or density of the coolant at these specific regions. The gamma detectors can also sense fission-product gas accumulation at the reactor head to advise of a failure of fuel-pin cladding.

  14. Material challenges for the next generation of fission reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Buckthorpe, Derek [AMEC, Knutsford, Cheshire (United Kingdom)


    The new generation of fission reactor systems wil require the deployment and construction of a series of advanced water cooled reactors as part of a package of measures to meet UK and European energy needs and to provide a near term non-fossil fuel power solution that addresses CO{sub 2} emission limits. In addition new longer term Generation IV reactor tye systems are being developed and evaluated to enhance safety, reliability, sustainability economics and proliferation resistance requirements and to meet alternative energy applications (outside of electricity generation) such as process heat and large scale hydrogen generation. New fission systems will impose significant challenges on materials supply and development. In the near term, because of the need to 'gear up' to large scale construction after decades of industrial hibernation/contraction and, in the longer term, because of the need for materials to operate under more challenging environments requiring the deployment and development of new alternative materials not yet established to an industrial stage. This paper investigates the materials challenges imposed by the new Generation III+ and Generation IV systems. These include supply and fabrication issues, development of new high temperature alloys and non-metallic materials, the use of new methods of manufacture and the best use of currently available resources and minerals. Recommendations are made as to how these materials challenges might be met and how governments, industry, manufacturers and researchers can all play their part. (orig.)

  15. Systems and methods for dismantling a nuclear reactor (United States)

    Heim, Robert R; Adams, Scott Ryan; Cole, Matthew Denver; Kirby, William E; Linnebur, Paul Damon


    Systems and methods for dismantling a nuclear reactor are described. In one aspect the system includes a remotely controlled heavy manipulator ("manipulator") operatively coupled to a support structure, and a control station in a non-contaminated portion of a workspace. The support structure provides the manipulator with top down access into a bioshield of a nuclear reactor. At least one computing device in the control station provides remote control to perform operations including: (a) dismantling, using the manipulator, a graphite moderator, concrete walls, and a ceiling of the bioshield, the manipulator being provided with automated access to all internal portions of the bioshield; (b) loading, using the manipulator, contaminated graphite blocks from the graphite core and other components from the bioshield into one or more waste containers; and (c) dispersing, using the manipulator, dust suppression and contamination fixing spray to contaminated matter.

  16. Development of fluid system design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. J.; Chang, M. H.; Kang, D. J. and others


    This study presents the technology development of the system design concepts of SMART, a multi-purposed integral reactor with enhanced safety and operability, for use in diverse usages and applications of the nuclear energy. This report contains the following; - Design characteristics - Performance and safety related design criteria - System description: Primary system, Secondary system, Residual heat removal system, Make-up system, Component cooling system, Safety system - Development of design computer code: Steam generator performance(ONCESG), Pressurizer performance(COLDPZR), Steam generator flow instability(SGINS) - Development of component module and modeling using MMS computer code - Design calculation: Steam generator thermal sizing, Analysis of feed-water temperature increase at a low flow rate, Evaluation of thermal efficiency in the secondary system, Inlet orifice throttling coefficient for the prevention of steam generator flow instability, Analysis of Nitrogen gas temperature in the pressurizer during heat-up process, evaluation of water chemistry and erosion etc. The results of this study can be utilized not only for the foundation technology of the next phase basic system design of the SMART but also for the basic model in optimizing the system concepts for future advanced reactors. (author)

  17. Designing visual displays and system models for safe reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.


    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  18. Subcritical flutter in the acoustics of friction

    National Research Council Canada - National Science Library

    O.N Kirillov


    ...-simple eigenfrequencies at the nodes. At contact with friction pads, the rotating continua, such as the singing wine glass or the squealing disc brake, start to vibrate owing to the subcritical flutter instability...

  19. Development of ROV System for FOSAR in Reactor Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Soo; Kim, Tae Won; Lee, Sung Uk; Jeong, Kyung Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Nam Kyun [Korea Plant Service and Engineering Co., Seongnam (Korea, Republic of)


    Foreign object in the reactor vessel is susceptible to damage the fuel. Prior to reloading fuel assemblies into the core, FOSAR(Foreign Object Search And Retrieval) activities were performed on and beneath the lower core plate with conventional equipment. However, the reactor vessel is limited to humans who are susceptible to radiation exposure, and conventional equipment is hard to access because of the complexity of the structure. To improve the convenience of use and retrieval ability in the under-core plate region, we are developing a FOSAR system carried by ROV (Remotely Operated Vehicle). In this paper, we describe a ROV system developed. The ROV system is composed of robot vehicle and remote control unit. The vehicle has 4 thrusters, tilt, camera, light and depth sensor, etc. Considering radiation damage, processors are not equipped on the vehicle. Control signals and sensing signals are transferred through umbilical cable. Remote control unit is composed of electric driving module and two computers which one is for the control and the other is for the detection of robot position. Control computer has a joystick user input and video/signal input, and transmit motor control signal and lens control signal via CAN/RS485 communication. And the other computers transmit information of vehicle position to the control computer via serial communication. Information of vehicle position is obtained through image processing algorithm. The acquiring camera of vehicle is on the flange of reactor vessel. Simulations on the detection of vehicle position are performed at the reactor vessel mockup which scaled down by 6 and verified to use in the control of robot by visual tracking. And functional test has been performed on the air condition. In the future, performance test will be carried out real sized mockup and underwater condition

  20. Complete degradation of Orange G by electrolysis in sub-critical water. (United States)

    Yuksel, Asli; Sasaki, Mitsuru; Goto, Motonobu


    Complete degradation of azo dye Orange G was studied using a 500 mL continuous flow reactor made of SUS 316 stainless steel. In this system, a titanium reactor wall acted as a cathode and a titanium plate-type electrode was used as an anode in a subcritical reaction medium. This hydrothermal electrolysis process provides an environmentally friendly route that does not use any organic solvents or catalysts to remove organic pollutants from wastewater. Reactions were carried out from 30 to 90 min residence times at a pressure of 7 MPa, and at different temperatures of 180-250°C by applying various direct currents ranging from 0.5 to 1A. Removal of dye from the product solution and conversion of TOC increased with increasing current value. Moreover, the effect of salt addition on degradation of Orange G and TOC conversion was investigated, because in real textile wastewater, many salts are also included together with dye. Addition of Na(2)CO(3) resulted in a massive degradation of the dye itself and complete mineralization of TOC, while NaCl and Na(2)SO(4) obstructed the removal of Orange G. Greater than 99% of Orange G was successfully removed from the product solution with a 98% TOC conversion. Copyright © 2011 Elsevier B.V. All rights reserved.

  1. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)


    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  2. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL


    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  3. Thermal hydraulic core simulation of the MYRRHA Reactor in steady state operation

    Energy Technology Data Exchange (ETDEWEB)

    Ferandes, Gustavo H.N.; Ramos, Mário C.; Carvalho, Athos M.S.S.; Cabrera, Carlos E.V.; Costa, Antonella L.; Pereira, Claubia, E-mail:, E-mail:, E-mail:, E-mail:, E-mail:, E-mail: [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)


    MYRRHA (Multi-purpose Hybrid Research Reactor for High-tech Applications) is a prototype nuclear subcritical reactor driven by a particle accelerator. As a special property, the reactor maintains the nuclear fission chain reaction by means of an external neutron source provided by a particle accelerator. The main aim of this work is to study two types of coolants, LBE (Lead-Bismuth Eutectic) and Na (Sodium) that are two strong candidates to be used in ADS systems as well as in Generation IV (GEN-IV) reactors. Firstly, it was developed a thermal hydraulic model of the MYRRHA core using the RELAP5-3D, considering LBE as coolant (original project). After this, the LBE was substituted by Na coolant to investigate the reactor behavior in such case. Results have demonstrated the high heat transfer capacity of the LBE coolant in this type of system. (author)

  4. Automated power control system for reactor TRIGA PUSPATI (United States)

    Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair


    Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

  5. Analysis of reactivity determination methods in the subcritical experiment Yalina (United States)

    Persson, Carl-Magnus; Seltborg, Per; Åhlander, Alexandra; Gudowski, Waclaw; Stummer, Thomas; Kiyavitskaya, Hanna; Bournos, Victor; Fokov, Yurij; Serafimovich, Ivan; Chigrinov, Sergey


    Different reactivity determination methods have been investigated, based on experiments performed at the subcritical assembly Yalina in Minsk, Belarus. The development of techniques for on-line monitoring of the reactivity level in a future accelerator-driven system (ADS) is of major importance for safe operation. Since an ADS is operating in a subcritical mode, the safety margin to criticality must be sufficiently large. The investigated methods are the Slope Fit Method, the Sjöstrand Method and the Source Jerk Method. The results are compared with Monte Carlo simulations performed with different nuclear data libraries. The results of the Slope Fit Method are in good agreement with the Monte Carlo simulation results, whereas the Sjöstrand Method appears to underestimate the criticality somewhat. The Source Jerk Method is subject to inadequate statistical accuracy.

  6. Effect of increasing nitrobenzene loading rates on the performance of anaerobic migrating blanket reactor and sequential anaerobic migrating blanket reactor/completely stirred tank reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Kuscu, Ozlem Selcuk [Department of Environmental Engineering, Faculty of Engineering and Architecture, Sueleyman Demirel University, 32360, Isparta (Turkey); Sponza, Delia Teresa, E-mail: [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, 35160, Izmir (Turkey)


    A laboratory scale anaerobic migrating blanket reactor (AMBR) reactor was operated at nitrobenzene (NB) loading rates increasing from 3.33 to 66.67 g NB/m{sup 3} day and at a constant hydraulic retention time (HRT) of 6 days to observe the effects of increasing NB concentrations on chemical oxygen demand (COD), NB removal efficiencies, bicarbonate alkalinity, volatile fatty acid (VFA) accumulation and methane gas percentage. Moreover, the effect of an aerobic completely stirred tank reactor (CSTR) reactor, following the anaerobic reactor, on treatment efficiencies was also investigated. Approximately 91-94% COD removal efficiencies were observed up to a NB loading rate of 30.00 g/m{sup 3} day in the AMBR reactor. The COD removal efficiencies decreased from 91% to 85% at a NB loading rate of 66.67 g/m{sup 3} day. NB removal efficiencies were approximately 100% at all NB loading rates. The maximum total gas, methane gas productions and methane percentage were found to be 4.1, 2.6 l/day and 59%, respectively, at a NB loading rate of 30.00 g/m{sup 3} day. The optimum pH values were found to be between 7.2 and 8.4 for maximum methanogenesis. The total volatile fatty acid (TVFA) concentrations in the effluent were 110 and 70 mg/l in the first and second compartments at NB loading rates as high as 66.67 and 6.67 g/m{sup 3} day, respectively, while they were measured as zero in the effluent of the AMBR reactor. In this study, from 180 mg/l NB 66 mg/l aniline was produced in the anaerobic reactor while aniline was completely removed and transformed to 2 mg/l of cathechol in the aerobic CSTR reactor. Overall COD removal efficiencies were found to be 95% and 99% for NB loading rates of 3.33 and 66.67 g/m{sup 3} day in the sequential anaerobic AMBR/aerobic CSTR reactor system, respectively. The toxicity tests performed with Photobacterium phosphoreum (LCK 480, LUMIStox) and Daphnia magna showed that the toxicity decreased with anaerobic/aerobic sequential reactor system from

  7. Ongoing Development of a Series Bosch Reactor System (United States)

    Abney, Morgan; Mansell, Matt; DuMez, Sam; Thomas, John; Cooper, Charlie; Long, David


    Future manned missions to deep space or planetary surfaces will undoubtedly require highly robust, efficient, and regenerable life support systems that require minimal consumables. To meet this requirement, NASA continues to explore a Bosch-based carbon dioxide reduction system to recover oxygen from CO2. In order to improve the equivalent system mass of Bosch systems, we seek to design and test a "Series Bosch" system in which two reactors in series are optimized for the two steps of the reaction, as well as to explore the use of in situ materials as carbon deposition catalysts. Here we report recent developments in this effort including assembly and initial testing of a Reverse Water-Gas Shift reactor (RWGSr) and initial testing of two gas separation membranes. The RWGSr was sized to reduce CO2 produced by a crew of four to carbon monoxide as the first stage in a Series Bosch system. The gas separation membranes, necessary to recycle unreacted hydrogen and CO2, were similarly sized. Additionally, we report results of preliminary experiments designed to determine the catalytic properties of Martian and Lunar regolith simulant for the carbon deposition step.

  8. The detector system of the Daya Bay reactor neutrino experiment (United States)

    An, F. P.; Bai, J. Z.; Balantekin, A. B.; Band, H. R.; Beavis, D.; Beriguete, W.; Bishai, M.; Blyth, S.; Brown, R. L.; Butorov, I.; Cao, D.; Cao, G. F.; Cao, J.; Carr, R.; Cen, W. R.; Chan, W. T.; Chan, Y. L.; Chang, J. F.; Chang, L. C.; Chang, Y.; Chasman, C.; Chen, H. Y.; Chen, H. S.; Chen, M. J.; Chen, Q. Y.; Chen, S. J.; Chen, S. M.; Chen, X. C.; Chen, X. H.; Chen, X. S.; Chen, Y. X.; Chen, Y.; Cheng, J. H.; Cheng, J.; Cheng, Y. P.; Cherwinka, J. J.; Chidzik, S.; Chow, K.; Chu, M. C.; Cummings, J. P.; de Arcos, J.; Deng, Z. Y.; Ding, X. F.; Ding, Y. Y.; Diwan, M. V.; Dong, L.; Dove, J.; Draeger, E.; Du, X. F.; Dwyer, D. A.; Edwards, W. R.; Ely, S. R.; Fang, S. D.; Fu, J. Y.; Fu, Z. W.; Ge, L. Q.; Ghazikhanian, V.; Gill, R.; Goett, J.; Gonchar, M.; Gong, G. H.; Gong, H.; Gornushkin, Y. A.; Grassi, M.; Greenler, L. S.; Gu, W. Q.; Guan, M. Y.; Guo, R. P.; Guo, X. H.; Hackenburg, R. W.; Hahn, R. L.; Han, R.; Hans, S.; He, M.; He, Q.; He, W. S.; Heeger, K. M.; Heng, Y. K.; Higuera, A.; Hinrichs, P.; Ho, T. H.; Hoff, M.; Hor, Y. K.; Hsiung, Y. B.; Hu, B. Z.; Hu, L. M.; Hu, L. J.; Hu, T.; Hu, W.; Huang, E. C.; Huang, H. Z.; Huang, H. X.; Huang, P. W.; Huang, X.; Huang, X. T.; Huber, P.; Hussain, G.; Isvan, Z.; Jaffe, D. E.; Jaffke, P.; Jen, K. L.; Jetter, S.; Ji, X. P.; Ji, X. L.; Jiang, H. J.; Jiang, W. Q.; Jiao, J. B.; Johnson, R. A.; Joseph, J.; Kang, L.; Kettell, S. H.; Kohn, S.; Kramer, M.; Kwan, K. K.; Kwok, M. W.; Kwok, T.; Lai, C. Y.; Lai, W. C.; Lai, W. H.; Langford, T. J.; Lau, K.; Lebanowski, L.; Lee, J.; Lee, M. K. P.; Lei, R. T.; Leitner, R.; Leung, J. K. C.; Lewis, C. A.; Li, B.; Li, C.; Li, D. J.; Li, F.; Li, G. S.; Li, J.; Li, N. Y.; Li, Q. J.; Li, S. F.; Li, S. C.; Li, W. D.; Li, X. B.; Li, X. N.; Li, X. Q.; Li, Y.; Li, Y. F.; Li, Z. B.; Liang, H.; Liang, J.; Lin, C. J.; Lin, G. L.; Lin, P. Y.; Lin, S. X.; Lin, S. K.; Lin, Y. C.; Ling, J. J.; Link, J. M.; Littenberg, L.; Littlejohn, B. R.; Liu, B. J.; Liu, C.; Liu, D. W.; Liu, H.; Liu, J. L.; Liu, J. C.; Liu, S.; Liu, S. S.; Liu, X.; Liu, Y. B.; Lu, C.; Lu, H. Q.; Lu, J. S.; Luk, A.; Luk, K. B.; Luo, T.; Luo, X. L.; Ma, L. H.; Ma, Q. M.; Ma, X. Y.; Ma, X. B.; Ma, Y. Q.; Mayes, B.; McDonald, K. T.; McFarlane, M. C.; McKeown, R. D.; Meng, Y.; Mitchell, I.; Mohapatra, D.; Monari Kebwaro, J.; Morgan, J. E.; Nakajima, Y.; Napolitano, J.; Naumov, D.; Naumova, E.; Newsom, C.; Ngai, H. Y.; Ngai, W. K.; Nie, Y. B.; Ning, Z.; Ochoa-Ricoux, J. P.; Olshevskiy, A.; Pagac, A.; Pan, H.-R.; Patton, S.; Pearson, C.; Pec, V.; Peng, J. C.; Piilonen, L. E.; Pinsky, L.; Pun, C. S. J.; Qi, F. Z.; Qi, M.; Qian, X.; Raper, N.; Ren, B.; Ren, J.; Rosero, R.; Roskovec, B.; Ruan, X. C.; Sands, W. R.; Seilhan, B.; Shao, B. B.; Shih, K.; Song, W. Y.; Steiner, H.; Stoler, P.; Stuart, M.; Sun, G. X.; Sun, J. L.; Tagg, N.; Tam, Y. H.; Tanaka, H. K.; Tang, W.; Tang, X.; Taychenachev, D.; Themann, H.; Torun, Y.; Trentalange, S.; Tsai, O.; Tsang, K. V.; Tsang, R. H. M.; Tull, C. E.; Tung, Y. C.; Viaux, N.; Viren, B.; Virostek, S.; Vorobel, V.; Wang, C. H.; Wang, L. S.; Wang, L. Y.; Wang, L. Z.; Wang, M.; Wang, N. Y.; Wang, R. G.; Wang, T.; Wang, W.; Wang, W. W.; Wang, X. T.; Wang, X.; Wang, Y. F.; Wang, Z.; Wang, Z.; Wang, Z. M.; Webber, D. M.; Wei, H. Y.; Wei, Y. D.; Wen, L. J.; Wenman, D. L.; Whisnant, K.; White, C. G.; Whitehead, L.; Whitten, C. A.; Wilhelmi, J.; Wise, T.; Wong, H. C.; Wong, H. L. H.; Wong, J.; Wong, S. C. F.; Worcester, E.; Wu, F. F.; Wu, Q.; Xia, D. M.; Xia, J. K.; Xiang, S. T.; Xiao, Q.; Xing, Z. Z.; Xu, G.; Xu, J. Y.; Xu, J. L.; Xu, J.; Xu, W.; Xu, Y.; Xue, T.; Yan, J.; Yang, C. G.; Yang, L.; Yang, M. S.; Yang, M. T.; Ye, M.; Yeh, M.; Yeh, Y. S.; Yip, K.; Young, B. L.; Yu, G. Y.; Yu, Z. Y.; Zeng, S.; Zhan, L.; Zhang, C.; Zhang, F. H.; Zhang, H. H.; Zhang, J. W.; Zhang, K.; Zhang, Q. X.; Zhang, Q. M.; Zhang, S. H.; Zhang, X. T.; Zhang, Y. C.; Zhang, Y. H.; Zhang, Y. M.; Zhang, Y. X.; Zhang, Y. M.; Zhang, Z. J.; Zhang, Z. Y.; Zhang, Z. P.; Zhao, J.; Zhao, Q. W.; Zhao, Y. F.; Zhao, Y. B.; Zheng, L.; Zhong, W. L.; Zhou, L.; Zhou, N.; Zhou, Z. Y.; Zhuang, H. L.; Zimmerman, S.; Zou, J. H.


    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of νbare oscillations over km-baselines. Subsequent data has provided the world's most precise measurement of sin2 2θ13 and the effective mass splitting Δ mee2. The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world's most prolific sources of electron antineutrinos. Multiple antineutrino detectors are deployed in three underground water pools at different distances from the reactor cores to search for deviations in the antineutrino rate and energy spectrum due to neutrino mixing. Instrumented with photomultiplier tubes, the water pools serve as shielding against natural radioactivity from the surrounding rock and provide efficient muon tagging. Arrays of resistive plate chambers over the top of each pool provide additional muon detection. The antineutrino detectors were specifically designed for measurements of the antineutrino flux with minimal systematic uncertainty. Relative detector efficiencies between the near and far detectors are known to better than 0.2%. With the unblinding of the final two detectors' baselines and target masses, a complete description and comparison of the eight antineutrino detectors can now be presented. This paper describes the Daya Bay detector systems, consisting of eight antineutrino detectors in three instrumented water pools in three underground halls, and their operation through the first year of eight detector data-taking.

  9. TCODE: a computer code for analysis of tritium and vacuum systems for tokamak fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Clemmer, R.G.


    TCODE can be used for either near-term experimental reactors or for commercial reactors. The code provides options for items that may be included in a commercial reactor such as a divertor, neutral beam heating, and a breeding blanket. The code was used to calculate tritium and vacuum system parameters for the near term reactors ITR, TNS-UP and EPR as well as for some commercial reactor designs, the UWMAK series. A selected sample of the tritium and vacuum parameters for these reactor designs is shown. Also shown are parameters for a hypothetical reactor UWMAK-III M having similar characteristics to UWMAK-III but with a higher fractional burnup (5.0% cf. 0.83%). The impact of the reactor design scenario upon major tritium and vacuum systems is discussed.

  10. Anaerobic sewage treatment in a one-stage UASB reactor and a combined UASB Digester system

    NARCIS (Netherlands)

    Mahmoud, N.A.; Zeeman, G.; Gijzen, H.J.; Lettinga, G.


    The treatment of sewage at 15°C was investigated in a one-stage upflow anaerobic sludge blanket (UASB) reactor and a UASB-Digester system. The latter consists of a UASB reactor complemented with a digester for mutual sewage treatment and sludge stabilisation. The UASB reactor was operated at a

  11. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip; Widi Setiawan [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)


    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  12. Designing a SCADA system simulator for fast breeder reactor (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.


    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  13. Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Amar, E-mail:; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P.S.; Bishnoi, Saroj


    Highlights: •Experimental subcritical facility BRAHMMA coupled to D-D/D-T neutron generator. •Preliminary results of PNS experiments reported. •Feynman-alpha noise measurements explored with continuous source. -- Abstract: The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated k{sub eff} of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k{sub s} and external neutron source efficiency φ{sup ∗} in great details. Experiments with D-T neutrons are also underway.

  14. Parametric systems analysis of the Modular Stellarator Reactor (MSR)

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.


    The close coupling in the stellarator/torsatron/heliotron (S/T/H) between coil design (peak field, current density, forces), magnetics topology (transform, shear, well depth), and plasma performance (equilibrium, stability, transport, beta) complicates the reactor assessment more so than for most magnetic confinement systems. In order to provide an additional degree of resolution of this problem for the Modular Stellarator Reactor (MSR), a parametric systems model has been developed and applied. This model reduces key issues associted ith plasma performance, first-wall/blanket/shield (FW/B/S), and coil design to a simple relationship between beta, system geometry, and a number of indicators of overall plant performance. The results of this analysis can then be used to guide more detailed, multidimensional plasma, magnetics, and coil design efforts towards technically and economically viable operating regimes. In general, it is shown that beta values > 0.08 may be needed if the MSR approach is to be substantially competitive with other approaches to magnetic fusion in terms of system power density, mass utilization, and cost for total power output around 4.0 GWt; lower powers will require even higher betas.

  15. Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

    Directory of Open Access Journals (Sweden)

    Isaac Skavdahl


    Full Text Available Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (Tco and the hot outlet temperature of the intermediate heat exchanger (Tho2 by manipulating the hot-side flow rates of the heat exchangers (Fh/Fh2 responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (Tco only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1 flow rate manipulation; (2 reactor power manipulation; or (3 a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

  16. Catalytic membrane reactor for tritium extraction system from He purge

    Energy Technology Data Exchange (ETDEWEB)

    Santucci, Alessia, E-mail: [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Incelli, Marco [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); DEIM, University of Tuscia, Via del Paradiso 47, 01100 Viterbo (Italy); Sansovini, Mirko; Tosti, Silvano [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy)


    Highlights: • In the HCBB blanket, the produced tritium is recovered by purging with helium; membrane technologies are able to separate tritium from helium. • The paper presents the results of two experimental campaigns. • In the first, a Pd–Ag diffuser for hydrogen separation is tested at several operating conditions. • In the second, the ability of a Pd–Ag membrane reactor for water decontamination is assessed by performing isotopic swamping and water gas shift reactions. - Abstract: In the Helium Cooled Pebble Bed (HCPB) blanket concept, the produced tritium is recovered purging the breeder with helium at low pressure, thus a tritium extraction system (TES) is foreseen to separate the produced tritium (which contains impurities like water) from the helium gas purge. Several R&D activities are running in parallel to experimentally identify most promising TES technologies: particularly, Pd-based membrane reactors (MR) are under investigation because of their large hydrogen selectivity, continuous operation capability, reliability and compactness. The construction and operation under DEMO relevant conditions (that presently foresee a He purge flow rate of about 10,000 Nm{sup 3}/h and a H{sub 2}/He ratio of 0.1%) of a medium scale MR is scheduled for next year, while presently preliminary experiments on a small scale reactor are performed to identify most suitable operative conditions and catalyst materials. This work presents the results of an experimental campaign carried out on a Pd-based membrane aimed at measuring the capability of this device in separating hydrogen from the helium. Many operative conditions have been investigated by considering different He/H{sub 2} feed flow ratios, several lumen pressures and reactor temperatures. Moreover, the performances of a membrane reactor (composed of a Pd–Ag tube having a wall thickness of about 113 μm, length 500 mm and diameter 10 mm) in processing the water contained in the purge gas have been

  17. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B.F.; Berry, D.T.


    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters.

  18. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)


    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  19. Utilisation of British University Research Reactors. (United States)

    Duncton, P. J.; And Others

    British experience relating to the employment of university research reactors and subcritical assemblies in the education of nuclear scientists and technologists, in the training of reactor operators and for fundamental pure and applied research in this field is reviewed. The facilities available in a number of British universities and the uses…

  20. Improved reactor regulating system logical architecture using genetic algorithm

    Directory of Open Access Journals (Sweden)

    Hyo-Sub Shim


    Full Text Available An improved Reactor Regulating System (RRS logic architecture, which is combined with genetic algorithm (GA, is implemented in this work. It is devised to provide an optimal solution to the current RRS. The current system works desirably and has contributed to safe and stable nuclear power plant operation. However, during the ascent and descent section of the reactor power, the RRS output reveals a relatively high steady-state error, and the output also carries a considerable level of overshoot. In an attempt to consolidate conservatism and minimize the error, this work proposes to apply GA to RRS and suggests reconfiguring the system. Prior to the use of GA, reverse engineering is implemented to build a Simulink-based RRS model. Reengineering is followed to produce a newly configured RRS to generate an output that has a reduced steady-state error and diminished overshoot level. A full-scope APR1400 simulator is used to examine the dynamic behaviors of RRS and to build the RRS Simulink model.

  1. Modification of reference temperature program in reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Sung Sik; Lee, Byung Jin; Kim, Se Chang; Cheong, Jong Sik [Korea Power Engineering Company, Inc., Seoul (Korea, Republic of); Kim, Ji In; Doo, Jin Yong [Korea Electric Power Cooperation, Yonggwang (Korea, Republic of)


    In Yonggwang nuclear units 3 and 4 currently under commercial operation, the cold temperature was very close to the technical specification limit of 298 deg C during initial startup testing, which was caused by the higher-than-expected reactor coolant system flow. Accordingly, the reference temperature (Tref) program needed to be revised to allow more flexibility for plant operations. In this study, the method of a specific test performed at Yonggwang nuclear unit 4 to revise the Tref program was described and the test results were discussed. In addition, the modified Tref program was evaluated on its potential impacts on system performance and safety. The methods of changing the Tref program and the associated pressurizer level setpoint program were also explained. Finally, for Ulchin nuclear unit 3 and 4 currently under initial startup testing, the effects of reactor coolant system flow rate on the coolant temperature were evaluated from the thermal hydraulic standpoint and an optimum Tref program was recommended. 6 refs., 4 figs., 2 tabs. (Author)

  2. Development of system integration technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Moon Hee; Kang, D. J.; Kim, K. K. and others


    The objective of this report is to integrate the conceptual design of an integral reactor, SMART producing thermal energy of 330 MW, which will be utilized to supply energy for seawater desalination and small-scale power generation. This project also aims to develop system integration technology for effective design of the reactor. For the conceptual design of SMART, preliminary design requirements including the top-tier requirements and design bases were evaluated and established. Furthermore, in the view of the application of codes and standards to the SMART design, existing laws, codes and standards were analyzed and evaluated with respect to its applicability. As a part of this evaluation, directions and guidelines were proposed for the development of new codes and standards which shall be applied to the SMART design. Regarding the integration of SMART conceptual designs, major design activities and interfaces between design departments were established and coordinated through the design process. For the effective management of all design schedules, a work performance evaluation system was developed and applied to the design process. As the results of this activity, an integrated output of SMART designs was produced. Two additional scopes performed in this project include the preliminary economic analysis on the SMART utilization for seawater desalination, and the planning of verification tests for technology implemented into SMART and establishing development plan of the computer codes to be used for SMART design in the next phase. The technical cooperation with foreign country and international organization for securing technologies for integral reactor design and its application was coordinated and managed through this project. (author)

  3. Development of small and medium integral reactor. ctor Development of fluid system design for small and medium integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. J.; Chang, M. H.; Kim, K. K.; Kim, J. P.; Yoon, J. H.; Lee, Y. J.; Park, C. T.; Bae, Y. Y.; Kang, D. J.; Lee, K. H.; Lee, J.; Kim, H. Y.; Cho, B. H.; Seo, J. K.; Kang, K. S.; Kang, H. O.


    The purpose of this study is to develop system design technology of integral reactor, as a new design concept of small and medium reactor having enhanced safety and economy, and to have a design assessment / verification technology through basic thermal hydraulic experiments. This report describes of the following: (1) basic requirement for the integral reactor system design (2) Conceptual design of primary and secondary circuits of NSSS, emergency core cooling system, passive residual heat removal system, severe accident mitigation cooling system, passive residual heat removal system, severe accident mitigation system and other auxiliary system (3) Requirements and test program for the basic thermal hydraulic experiments including, CHF test for hexagonal fuel assembly, flow instability for once-through steam generator, core flow distribution test and verification test for non-condensable gas model in RELAP-5 code. The results of this study can be utilized for using as the foundation technology of in the next basic design phase and design technology for future advanced reactors. (author). 30 refs.,24 tabs., 56 figs.

  4. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)


    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  5. Implementation of a management system for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo, E-mail: [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes, E-mail: araquino@ipen.b, E-mail: dmzouain@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)


    This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

  6. High Flux Isotope Reactor system RELAP5 input model

    Energy Technology Data Exchange (ETDEWEB)

    Morris, D.G.; Wendel, M.W.


    A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model.

  7. The MAUS nuclear space reactor with ion propulsion system (United States)

    Mainardi, Enrico


    MAUS (Moltiplicatore Avanzato Ultracompatto Spaziale) is a nuclear reactor concept design capable to ensure a reliable, long-lasting, low-mass, compact energy supply needed for advanced, future space missions. The exploration of the solar system and the space beyond requires the development of nuclear energy generators for supplying electricity to space-bases, spacecrafts, probes or satellites, as well as for propelling ships in long space missions. For propulsion, the MAUS nuclear reactor could be used to power electric ion drive engines. An ion engine is able to build up to very high velocities, far greater than chemical propulsion systems, but has high power and long service requirements. The MAUS concept is described, together with the ion propulsion engine and together with the reference thermoionic process used to convert the thermal power into electricity. The design work has been performed at the Nuclear Engineering and Energy Conversion Department of the University of Rome "La Sapienza" starting from 1992 on an issue submitted by the Italian Space Agency (ASI), in cooperation with the research laboratories of ENEA.

  8. Criticality Safety Evaluation of the LLNL Inherently Safe Subcritical Assembly (ISSA)

    Energy Technology Data Exchange (ETDEWEB)

    Percher, Catherine [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)


    The LLNL Nuclear Criticality Safety Division has developed a training center to illustrate criticality safety and reactor physics concepts through hands-on experimental training. The experimental assembly, the Inherently Safe Subcritical Assembly (ISSA), uses surplus highly enriched research reactor fuel configured in a water tank. The training activities will be conducted by LLNL following the requirements of an Integration Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of LLNL instructors. This report provides the technical criticality safety basis for instructional operations with the ISSA experimental assembly.

  9. Modeling of Parameters of Subcritical Assembly SAD

    CERN Document Server

    Petrochenkov, S; Puzynin, I


    The accepted conceptual design of the experimental Subcritical Assembly in Dubna (SAD) is based on the MOX core with a nominal unit capacity of 25 kW (thermal). This corresponds to the multiplication coefficient $k_{\\rm eff} =0.95$ and accelerator beam power 1 kW. A subcritical assembly driven with the existing 660 MeV proton accelerator at the Joint Institute for Nuclear Research has been modelled in order to make choice of the optimal parameters for the future experiments. The Monte Carlo method was used to simulate neutron spectra, energy deposition and doses calculations. Some of the calculation results are presented in the paper.

  10. Mechanisms of Subcritical Cracking in Calcite (United States)

    Royne, A.; Dysthe, D. K.; Bisschop, J.


    Brittle materials are characterized by a critical stress intensity factor above which they will fail catastrophically by dynamic cracking. However, it has been observed that materials can also fail at much lower stresses, through slow crack growth, often referred to as subcritical cracking. This phenomenon can take place even in vacuum, but is greatly enhanced by water and other reactive species in the environment. For a given material and environmental condition there is a systematic relationship between the crack tip velocity and the stress intensity factor. The presence of a lower stress limit to subcritical cracking has been predicted from thermodynamics but has not been firmly demonstrated experimentally. This parameter would control the long- term strength of geological materials. Subcritical cracking must necessarily be important in controlling the rock strength in near-surface processes where water and other active species are present and the displacements and stresses are low. Weathering is one example of such a process. Modelling has shown that fracture networks generated by a high degree of subcritical cracking will percolate at much lower fracture densities than purely stochastical fracture networks. This has important implications for how water can move through the crust. Understanding the mechanisms for subcritical crack growth in geological materials is also important in assessing the stability and long term performance of sequestration reservoirs for CO2 or nuclear waste. The mechanism for stress corrosion is well known for glasses and quartz. For carbonate minerals, the mechanism for subcritical crack growth has not been identified, and the only experimental studies on calcitic materials have been on polycrystalline rocks such as marble. Suggested mechanisms include stress corrosion (weakening reactions at the crack tip), preferential dissolution at the crack tip with rapid removal of dissolved species, and environmentally controlled

  11. Reactor/Brayton power systems for nuclear electric spacecraft (United States)

    Layton, J. P.


    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  12. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)


    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  13. Reliability of digital reactor protection system based on extenics. (United States)

    Zhao, Jing; He, Ya-Nan; Gu, Peng-Fei; Chen, Wei-Hua; Gao, Feng


    After the Fukushima nuclear accident, safety of nuclear power plants (NPPs) is widespread concerned. The reliability of reactor protection system (RPS) is directly related to the safety of NPPs, however, it is difficult to accurately evaluate the reliability of digital RPS. The method is based on estimating probability has some uncertainties, which can not reflect the reliability status of RPS dynamically and support the maintenance and troubleshooting. In this paper, the reliability quantitative analysis method based on extenics is proposed for the digital RPS (safety-critical), by which the relationship between the reliability and response time of RPS is constructed. The reliability of the RPS for CPR1000 NPP is modeled and analyzed by the proposed method as an example. The results show that the proposed method is capable to estimate the RPS reliability effectively and provide support to maintenance and troubleshooting of digital RPS system.

  14. N-reactor charge-discharge system analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tokarz, R.D.; Marr, G.D.; Nesbitt, J.F.


    This report documents an analysis of the existing systems in the N-Reactor fuel flow path. It recommends equipment improvements and changes in that path to allow the charge-discharge rates to be increased to 500 tubes per outage without increasing reactor outage time. The estimated program cost of $14 million is projected over an estimated 3-year period. It does not include costs detailed as part of the existing restoration program or any costs that are considered as normal maintenance. The recommendations contained in this report provide a direction and goal for every critical aspect of the fuel flow path. The way in which these recommendations are implemented may greatly affect the schedule and costs. Previous studies by UNC have shown that enhanced fuel element handling has the potential of increasing productivity by 33 days at a cost benefit estimated at $18 million per year. Enhanced fuel handling provides the greatest potential for productivity improvement of any of the areas considered in these studies.


    Energy Technology Data Exchange (ETDEWEB)



    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements. The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprising two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with ''green'' (fresh) fuel and one with spent fuel. Both the green and spent fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, 3 green fuel and 4 spent fuel loading configurations were considered to serve as benchmark models. However, shortcomings in experimental data failed to meet the high standards for a benchmark problem. Nevertheless, the data provided by these subcritical measurements can

  16. Supervisory Control System Architecture for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cetiner, Sacit M [ORNL; Cole, Daniel L [University of Pittsburgh; Fugate, David L [ORNL; Kisner, Roger A [ORNL; Melin, Alexander M [ORNL; Muhlheim, Michael David [ORNL; Rao, Nageswara S [ORNL; Wood, Richard Thomas [ORNL


    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  17. Compatibility of refractory materials for nuclear reactor poison control systems (United States)

    Sinclair, J. H.


    Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.

  18. Nuclear plant-aging research on reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.C.


    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  19. Testing of an advanced thermochemical conversion reactor system

    Energy Technology Data Exchange (ETDEWEB)


    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  20. Numerical simulation of the power characteristics of twin-core pulse reactor-pumped laser system (United States)

    Gulevich, A. V.; Barzilov, A. P.; Dyachenko, P. P.; Zrodnikov, A. V.; Kukharchuk, O. F.; Kachanov, B. V.; Kolyada, S. G.; Pashin, E. A.


    Concept for high-power pulsed reactor-pumped laser system (RPLS) based on the new physical principles (direct nuclear-to-optical conversion) is discussed with reference to ICF feasibility problem. Theoretical problems for substantiation of the neutronic and physical characteristics of the RPLS power model are considered. Results of numerical studies of the expected power characteristics of reactor laser system are discussed.

  1. Systems and methods for enhancing isolation of high-temperature reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per F.


    A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

  2. Subcritical water extraction of flavoring and phenolic compounds from cinnamon bark (Cinnamomum zeylanicum). (United States)

    Khuwijitjaru, Pramote; Sayputikasikorn, Nucha; Samuhasaneetoo, Suched; Penroj, Parinda; Siriwongwilaichat, Prasong; Adachi, Shuji


    Cinnamon bark (Cinnamomum zeylanicum) powder was treated with subcritical water at 150 and 200°C in a semi-continuous system at a constant flow rate (3 mL/min) and pressure (6 MPa). Major flavoring compounds, i.e., cinnamaldehyde, cinnamic acid, cinnamyl alcohol and coumarin, were extracted at lower recoveries than the extraction using methanol, suggesting that degradation of these components might occur during the subcritical water treatment. Caffeic, ferulic, p-coumaric, protocatechuic and vanillic acids were identified from the subcritical water treatment. Extraction using subcritical water was more effective to obtain these acids than methanol (50% v/v) in both number of components and recovery, especially at 200°C. Subcritical water treatment at 200°C also resulted in a higher total phenolic content and DPPH radical scavenging activity than the methanol extraction. The DPPH radical scavenging activity and total phenolic content linearly correlated but the results suggested that the extraction at 200°C might result in other products that possessed a free radical scavenging activity other than the phenolic compounds.

  3. Determination of subcriticality and effective source strength by source drop and jerk experiments

    Energy Technology Data Exchange (ETDEWEB)

    Taninaka, Hiroshi [Interdisciplinary Graduate School of Science and Technology, Kinki University, 3-4-1, Kowakae, Higashi-Osaka, 577-8502 (Japan); Hashimoto, Kengo [Atomic Energy Research Institute, Kinki University, 3-4-1, Kowakae, Higashi-Osaka, 577-8502 (Japan)


    This paper presents applicability of least squares inverse kinetics method (LSIKM) to source drop and source jerk experiments. The LSIKM can estimate both reactivity and source strength by applying least square approximation to a correlation between time-sequence count data and inverse kinetics analysis data. The experiments were performed in the UTR-KINKI reactor to demonstrate the applicability of the LSIKM. To source jerk data, for comparison, conventional integral method is also applied. In the subcriticality and source strength obtained by the LSIKM, spatial dependence is slightly observed. However, the integral method leads to significant spatial dependence. The sub-criticalities inferred from source drop data are consistent with the results from source jerk data. (authors)

  4. Influence of moderator to fuel ratio (MFR) on burning thorium in a subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Wojciechowski, Andrzej, E-mail: [National Center for Nuclear Research, Otwock-Swierk (Poland); Joint Institute for Nuclear Research, Dubna (Russian Federation)


    The conversion ratio (CR) of Th-232 to U-233 calculation results for a subcritical reactor assembly is presented as a function of MFR, burnup, power density (PD) and fissile concentration. The calculated model is based on subcritical assembly which makes configuration of fuel rods and volumes of moderator and coolant changes possible. This comfortable assembly enables investigation of CR in a thorium cycle for different value of MFR. Additionally, the calculation results of U-233 saturation concentration are explained by mathematical model. The value of MFR main influences the saturation concentration of U-233 and fissile and the fissile concentration dependence of CR. The saturation value of CR is included in the range CR ∈ (0.911, 0.966) and is a slowly increasing function of MFR. The calculations were done with a MCNPX 2.7 code.

  5. Coupling of subcritical methanol with acidic ionic liquids for the acidity reduction of naphthenic acids

    Directory of Open Access Journals (Sweden)

    Zafar Faisal


    Full Text Available The presence of naphthenic acids (NAs in crude oil is the major cause of corrosion in the refineries and its processing equipment. The goal of this study is to reduce the total acid number (TAN of NAs by treating them with subcritical methanol in the presence of acidic ionic liquid (AIL catalysts. Experiments were carried out in an autoclave batch reactor and the effect of different reaction parameters was investigated. It was observed that TAN reduction was positively dependent on the temperature and concentration of the AIL whereas excess of methanol has a negative effect. Approximately 90% TAN reduction was achieved under the optimized reaction conditions using [BMIM]HSO4 as catalyst. It was also perceived from the experimental results that the AILs with longer alkyl chain exhibited higher catalytic activity. The activity and stability of AIL showed that they can be promising catalyst to esterify NAs under subcritical methanol.

  6. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip


    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  7. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A


    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  8. Dynamical Safety Analysis of the SABR Fusion-Fission Hybrid Reactor (United States)

    Sumner, Tyler; Stacey, Weston; Ghiaassian, Seyed


    A hybrid fusion-fission reactor for the transmutation of spent nuclear fuel is being developed at Georgia Tech. The Subcritical Advanced Burner Reactor (SABR) is a 3000 MWth sodium-cooled, metal TRU-Zr fueled fast reactor driven by a tokamak fusion neutron source based on ITER physics and technology. We are investigating the accident dynamics of SABR's coupled fission, fusion and heat removal systems to explore the safety characteristics of a hybrid reactor. Possible accident scenarios such as loss of coolant mass flow (LOFA), of power (LOPA) and of heat sink (LOHSA), as well as inadvertent reactivity insertions and fusion source excursion are being analyzed using the RELAP5-3D code, the ATHENA version of which includes liquid metal coolants.

  9. Reactor safeguards

    CERN Document Server

    Russell, Charles R


    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  10. Shielding considerations for advanced space nuclear reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Angelo, J.P. Jr.; Buden, D.


    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  11. Subcritical measurements of the WINCO slab tank experiment using the source-jerk technique

    Energy Technology Data Exchange (ETDEWEB)

    Spriggs, G.D.; Hansen, G.E.; Martin, E.R.; Plassmann, E.A.; Pederson, R.A.; Schlesser, J.A.; Krawczyk, T.L.; Tanner, J.E.; Smolen, G.R. (Los Alamos National Lab., NM (USA); Martin Marietta Energy Systems, Inc., Oak Ridge, TN (USA); Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (USA); Martin Marietta Energy Systems, Inc., Oak Ridge, TN (USA))


    Subcritical measurements of the WINCO slab tank using the source-jerk technique are presented. This technique determines subcriticality by analyzing the transient response produced by the sudden removal of an extraneous neutron source (i.e., a source jerk). We have found that the technique can provide an accurate means of measuring k in configurations that are close to critical (i.e., 0.90 < k < 1.0). As the system becomes more subcritical (i.e., k < 0.90), spatial effects introduce significant biases depending on the source and detector positions. A comparison between the measurements and Monte Carlo code calculations is also presented. 15 refs., 6 figs., 2 tabs.


    Directory of Open Access Journals (Sweden)



    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

  13. The fluidized bed reactor with a prepolymerization system and its influence on polymer physicochemical characteristics

    Directory of Open Access Journals (Sweden)

    Fernandes F.A.N.


    Full Text Available This work addresses the influence of a prepolymerization system on the behavior of the fluidized bed reactor used for polyethylene production. Its influence on the polymer's physicochemical characteristics and production was also studied. The results indicate that the use of prepolymerized catalyst particles results in milder temperatures in the fluidized bed reactor, thus avoiding the formation of hot spots, melting of the polymer particle and reactor shutdown. Productivity can be enhanced depending on the operational conditions used in the prepolymerization reactor.

  14. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin


    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  15. BEACON TSM application system to the operation of PWR reactors; Aplicacion del Sistema BEACON TSM a la operacion de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

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


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

  16. Pressurized hydrogenotrophic denitrification reactor for small water systems. (United States)

    Epsztein, Razi; Beliavski, Michael; Tarre, Sheldon; Green, Michal


    The implementation of hydrogenotrophic denitrification is limited due to safety concerns, poor H2 utilization and low solubility of H2 gas with the resulting low transfer rate. The current paper presents the main research work conducted on a pressurized hydrogenotrophic reactor for denitrification that was recently developed. The reactor is based on a new concept suggesting that a gas-liquid equilibrium is achieved in the closed headspace of denitrifying reactor, further produced N2 gas is carried out by the effluent and gas purging is not required. The feasibility of the proposed reactor was shown for two effluent concentrations of 10 and 1 mg NO3--N/L. Hydrogen gas utilization efficiencies of 92.8% and 96.9% were measured for the two effluent concentrations, respectively. Reactor modeling predicted high denitrification rates above 4 g NO3--N/(Lreactor·d) at reasonable operational conditions. Hydrogen utilization efficiency was improved up to almost 100% by combining the pressurized reactor with a following open-to-atmosphere polishing unit. Also, the potential of the reactor to remove ClO4- was shown. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Theoretic programs related to spallation neutron source for ADS. Pt.2: Thick target simulations The ADS is stated for Accelector Driven System

    CERN Document Server

    Fan Sheng; Shen Qing Biao; Zhao Zhi Xiang


    The spallation neutron target for intermediate energy proton incident is an important link for accelerator and subcritical reactor of accelerator driven system (ADS). The theoretic programs and Monte-Carlo codes are a useful approach for solving the physics of spallation neutron source. The authors discuss those codes at present work and introduce the application and development of SHIELD code

  18. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira


    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  19. Decay heat measurement on fusion reactor materials and validation of calculation code system

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment


    Decay heat rates for 32 fusion reactor relevant materials irradiated with 14-MeV neutrons were measured for the cooling time period between 1 minute and 400 days. With using the experimental data base, validity of decay heat calculation systems for fusion reactors were investigated. (author)

  20. A small, 1400 deg Kelvin, reactor for Brayton space power systems (United States)

    Lantz, E.; Mayo, W.


    A preliminary cost estimate for a small reactor in Brayton space power systems with (u-233)n or (pu-239)n as the fuel in the T-111 fuel elements totaled to about four million dollars; considered is a 22.8 in. diameter reactor with 247 fuel elements.

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


    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.'' DG-1277...

  2. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors (United States)


    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power reactors...

  3. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor (United States)

    Rohanda, Anis; Waris, Abdul


    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on 16O(n,p)16N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  4. Kinetic Parameter Measurements in the MINERVE Reactor (United States)

    Perret, Grégory; Geslot, Benoit; Gruel, Adrien; Blaise, Patrick; Di-Salvo, Jacques; De Izarra, Grégoire; Jammes, Christian; Hursin, Mathieu; Pautz, Andréas


    In the framework of an international collaboration, teams of the PSI and CEA research institutes measure the critical decay constant (α0 = β/A), delayed neutron fraction (β) and generation time (A) of the Minerve reactor using the Feynman-α, Power Spectral Density and Rossi-α neutron noise measurement techniques. These measurements contribute to the experimental database of kinetic parameters used to improve nuclear data files and validate modern methods in Monte Carlo codes. Minerve is a zero-power pool reactor composed of a central experimental test lattice surrounded by a large aluminum buffer and four high-enriched driver regions. Measurements are performed in three slightly subcritical configurations (-2 cents to -30 cents) using two high-efficiency 235U fission chambers in the driver regions. Measurement of α0 and β obtained by the two institutes and with the different techniques are consistent for the configurations envisaged. Slight increases of the β values are observed with the subcriticality level. Best estimate values are obtained with the Cross-Power Spectral Density technique at -2 cents, and are worth: β = 716.9±9.0 pcm, α0 = 79.0±0.6 s-1 and A = 90.7±1.4 μs. The kinetic parameters are predicted with MCNP5-v1.6 and TRIPOLI4.9 and the JEFF-3.1/3.1.1 and ENDF/B-VII.1 nuclear data libraries. The predictions for β and α0 overestimate the experimental results by 3-5% and 10-12%, respectively; that for A underestimate the experimental result by 6-7%. The discrepancies are suspected to come from the driven system nature of Minerve and the location of the detectors in the driver regions, which prevent accounting for the full reactor.

  5. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess


    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

  6. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system (United States)

    Jefferies, K. S.; Tew, R. C.


    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  7. Thermal power measurement based on Feynman-alpha correlation analysis in a low-power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Atsuko; Taninaka, Hiroshi [Interdisciplinary Graduate School of Science and Technology, Kinki University, 3-4-1, Kowakae, Higashi-Osaka, 577-8502 (Japan); Hashimoto, Kengo [Atomic Energy Research Institute, Kinki University, 3-4-1, Kowakae, Higashi-Osaka, 577-8502 (Japan)


    This paper presents applicability of the extended Feynman-a correlation method to reactor power measurement. In the extended method, higher-order difference filters are implemented and dead-time effect of neutron counter is considered. A series of the correlation measurements were performed in the UTR-KINKI reactor to demonstrate the applicability of the extended method. At a critical state, the reactor power inferred from saturated correlation amplitude is consistent with indication of linear power monitor of the reactor. At subcritical states, not only the correlation amplitudes but also the subcriticality of these states require for the determination of reactor power. In prompt decay constants and sub-criticalities obtained from the constants, detector-position dependence, i.e., spatial effect has significantly observed. These sub-criticalities have also led to the significant spatial dependence of the reactor power inferred. When reference sub-criticalities determined from source jerk experiment have employed instead of the spatially dependent sub-criticalities, the inferred reactor power has slight spatial dependence and agrees with indication of linear power monitor. (authors)

  8. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mazidi, S., E-mail: [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)


    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

  9. Investigation of Anaerobic Fluidized Bed Reactor/ Aerobic Moving Bed Bio Reactor (AFBR/MMBR) System for Treatment of Currant Wastewater (United States)

    JAFARI, Jalil; MESDAGHINIA, Alireza; NABIZADEH, Ramin; FARROKHI, Mehrdad; MAHVI, Amir Hossein


    Background: Anaerobic treatment methods are more suitable for the treatment of concentrated wastewater streams, offer lower operating costs, the production of usable biogas product. The aim of this study was to investigate the performance of an Anaerobic Fluidized Bed Reactor (AFBR)-Aerobic Moving Bed Bio Reactor (MBBR) in series arrangement to treat Currant wastewater. Methods: The bed materials of AFBR were cylindrical particles made of PVC with a diameter of 2–2.3 mm, particle density of 1250 kg/m3. The volume of all bed materials was 1.7 liter which expanded to 2.46 liters in fluidized situation. In MBBR, support media was composed of 1.5 liters Bee-Cell 2000 having porosity of 87% and specific surface area of 650m2/m3. Results: When system operated at 35 ºC, chemical oxygen demand (COD) removal efficiencies were achieved to 98% and 81.6% for organic loading rates (OLR) of 9.4 and 24.2 g COD/l.d, and hydraulic retention times (HRT) of 48 and 18 h, in average COD concentration feeding of 18.4 g/l, respectively. Conclusion: The contribution of AFBR in total COD removal efficiency at an organic loading rate (OLR) of 9.4 g COD/l.d was 95%, and gradually decreased to 76.5% in OLR of 24.2 g COD/l.d. Also with increasing in organic loading rate the contribution of aerobic reactor in removing COD gradually decreased. In this system, the anaerobic reactor played the most important role in the removal of COD, and the aerobic MBBR was actually needed to polish the anaerobic treated wastewater. PMID:26056640

  10. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V


    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  11. Investigating the breeding capabilities of hybrid soliton reactors

    Energy Technology Data Exchange (ETDEWEB)

    Catsaros, N., E-mail: [Institute of Nuclear and Radiological Sciences and Technology, Energy and Safety, National Centre for Scientific Research “Demokritos”, 27, Neapoleos Str., 15341 Aghia Paraskevi (Greece); Gaveau, B., E-mail: [Université Pierre et Marie Curie, Campus Jussieu, 75252 Paris Cedex 05 (France); Jaekel, M.-T., E-mail: [Laboratoire de Physique Théorique de l’Ecole Normale Supérieure (CNRS), 24 rue Lhomond, 75231 Paris Cedex 05 (France); Jejcic, A. [Laboratoire de Physique Théorique de l’Ecole Normale Supérieure (CNRS), 24 rue Lhomond, 75231 Paris Cedex 05 (France); Maillard, J., E-mail: [Institut National de Physique Nucléaire et de Physique des Particules (CNRS), 3 rue Michel Ange, 75794 Paris Cedex 16 (France); Institut du Développement et des Ressources en Informatique Scientifique (CNRS), Campus Universitaire d’Orsay, rue John Von Neumann, Bat 506, 91403 Orsay Cedex (France); Maurel, G., E-mail: [Université Pierre et Marie Curie, Campus Jussieu, 75252 Paris Cedex 05 (France); Savva, P., E-mail: [Institute of Nuclear and Radiological Sciences and Technology, Energy and Safety, National Centre for Scientific Research “Demokritos”, 27, Neapoleos Str., 15341 Aghia Paraskevi (Greece); Silva, J., E-mail: [Université Pierre et Marie Curie, Campus Jussieu, 75252 Paris Cedex 05 (France); and others


    Highlights: • ANET code simulates innovative reactor designs including Accelerator Driven Systems. • Preliminary analysis of thermal hybrid soliton reactor examines breeding capabilities. • Subsequent studies will aim at optimizing parameters examined in this analysis. • Breeding capacity could be obtained while preserving efficiency and reactor stability. -- Abstract: Nuclear energy industry asks for an optimized exploitation of available natural resources and a safe operation of reactors. A closed fuel cycle requires the mass of fissile material depleted in a reactor to be equal to or less than the fissile mass produced in the same or in other reactors. In this work, a simple closed cycle scheme is investigated, grounded on the use of a conceptual thermal water-cooled and moderated subcritical hybrid soliton reactor (HSR). The concept is a specific Accelerator Driven System (ADS) operating at lower power than usual pressurized water reactors (PWRs). This type of reactor can be inherently safe, since shutdown is achieved by simply interrupting the accelerator's power supply. In this work a preliminary investigation is attempted concerning the existence of conditions under which the operation of a thermal HSR in breeding regime is possible. For this purpose, a conceptual encapsulated core has been defined by choosing the magnitude of a set of parameters which are important from the neutronic point of view, such as core geometry and fuel composition. Indications of breeding operation regime for thermal HSR systems are sought by performing preliminary simulations of this core. For this purpose, the Monte Carlo code ANET, which is being developed based on the high energy physics code GEANT is utilized, as being capable of simulating particles’ transport and interactions produced, including also simulation of low energy neutrons transport. A simple analytical model is also developed and presented in order to investigate the conditions under which

  12. Design and installation of a hot water layer system at the Tehran research reactor

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila


    Full Text Available A hot water layer system (HWLS is a novel system for reducing radioactivity under research reactor containment. This system is particularly useful in pool-type research reactors or other light water reactors with an open pool surface. The main purpose of a HWLS is to provide more protection for operators and reactor personnel against undesired doses due to the radio- activity of the primary loop. This radioactivity originates mainly from the induced radioactivity contained within the cooling water or probable minute leaks of fuel elements. More importantly, the bothersome radioactivity is progressively proportional to reactor power and, thus, the HWLS is a partial solution for mitigating such problems when power upgrading is planned. Following a series of tests and checks for different parameters, a HWLS has been built and put into operation at the Tehran research reactor in 2009. It underwent a series of comprehensive tests for a period of 6 months. Within this time-frame, it was realized that the HWLS could provide a better protection for reactor personnel against prevailing radiation under containment. The system is especially suitable in cases of abnormality, e. g. the spread of fission products due to fuel failure, because it prevents the mixing of pollutants developed deep in the pool with the upper layer and thus mitigates widespread leakage of radioactivity.

  13. Application of neutron activation analysis system in Xi'an pulsed reactor

    CERN Document Server

    Zhang Wen Shou; Yu Qi


    Neutron Activation Analysis System in Xi'an Pulsed Reactor is consist of rabbit fast radiation system and experiment measurement system. The functions of neutron activation analysis are introduced. Based on the radiation system. A set of automatic data handling and experiment simulating system are built. The reliability of data handling and experiment simulating system had been verified by experiment

  14. Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor (United States)

    Heinold, Mark R.; Berger, John F.; Loper, Milton H.; Runkle, Gary A.


    Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

  15. Subcritical experiments at the FREYA experiment; Experimentos subcriticos en el proyecto FREYA

    Energy Technology Data Exchange (ETDEWEB)

    Becares Palacios, V.; Villamarin fernandez, D.


    The FREYA Project of the 7th Framework Program is aimed to the study of the kinetics of subcritical reactors coupled to an external neutron source, and, more specifically, to the validation of reactivity monitoring techniques. CIEMAT activities within the frame of this project have consisted in analyzing the possible ways of correcting the spatial and energy effects on these reactivity monitoring techniques, as well as analyzing the effects that may have on them the presence of different materials in the reflector and the position of the neutron source.

  16. Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)


    This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

  17. Modelling of Mass Transfer Phenomena in Chemical and Biochemical Reactor Systems using Computational Fluid Dynamics

    DEFF Research Database (Denmark)

    Larsson, Hilde Kristina

    the velocity and pressure distributions in a fluid. CFD also enables the modelling of several fluids simultaneously, e.g. gas bubbles in a liquid, as well as the presence of turbulence and dissolved chemicals in a fluid, and many other phenomena. This makes CFD an appreciated tool for studying flow structures......, mixing, and other mass transfer phenomena in chemical and biochemical reactor systems. In this project, four selected case studies are investigated in order to explore the capabilities of CFD. The selected cases are a 1 ml stirred microbioreactor, an 8 ml magnetically stirred reactor, a Rushton impeller...... stirred pilot plant reactor, and a rotating bed reactor filled with catalytic porous material. A selection of the simulated phenomena includes the velocities and turbulent quantities in the reactors, as well as the distribution of the gas and liquid phases in them. Mixing times, oxygen transfer rates...

  18. Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Song, C. H.; Chung, M. K.; Park, C. K. and others


    The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

  19. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Nobuyuki; Suzuki, Katsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment


    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  20. System modeling for the advanced thermionic initiative single cell thermionic space nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H.H.; Lewis, B.R.; Klein, A.C. (Department of Nuclear Engineering, Oregon State University, Radiation Center, C116, Corvallis, Oregon 97331-5902 (United States)); Pawlowski, R.A. (Battelle Pacific Northwest Laboratories, Richland, Washington 99352 (United States))


    Incore thermionic space reactor design concepts which operate in a nominal power output range of 20 to 40 kWe are described. Details of the neutronics, thermionic, shielding, and heat rejection performance are presented. Two different designs, ATI-Driven and ATI-Driverless, are considered. Comparison of the core overall performance of these two configurations are described. The comparison of these two cores includes the overall conversion efficiency, reactor mass, shield mass, and heat rejection mass. An overall system design has been developed to model the advanced incore thermionic energy conversion based nuclear reactor systems for space applications in this power range.

  1. High Efficiency Microchannel Sabatier Reactor System for In Situ Resource Utilization Project (United States)

    National Aeronautics and Space Administration — An innovative Microchannel Sabatier Reactor System (MSRS) is proposed for 100% recovery of oxygen (as water) and methane from carbon dioxide (CO2), a valuable in...

  2. Modeling, simulation, and optimization of a front-end system for acetylene hydrogenation reactors

    Directory of Open Access Journals (Sweden)

    R. Gobbo


    Full Text Available The modeling, simulation, and dynamic optimization of an industrial reaction system for acetylene hydrogenation are discussed in the present work. The process consists of three adiabatic fixed-bed reactors, in series, with interstage cooling. These reactors are located after the compression and the caustic scrubbing sections of an ethylene plant, characterizing a front-end system; in contrast to the tail-end system where the reactors are placed after the de-ethanizer unit. The acetylene conversion and selectivity profiles for the reactors are optimized, taking into account catalyst deactivation and process constraints. A dynamic optimal temperature profile that maximizes ethylene production and meets product specifications is obtained by controlling the feed and intercoolers temperatures. An industrial acetylene hydrogenation system is used to provide the necessary data to adjust kinetics and transport parameters and to validate the approach.

  3. A pragmatic approach towards designing a second shutdown system for Tehran research reactor

    National Research Council Canada - National Science Library

    Boustani Ehsan; Khakshournia Samad; Khalafi Hossein


    One second shutdown system is proposed for the Tehran Research Reactor to achieve the goal of higher safety in compliance with current operational requirements and regulations and improve the overall...

  4. Progress in space nuclear reactor power systems technology development - The SP-100 program (United States)

    Davis, H. S.


    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

  5. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L


    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  6. The detector system of the Daya Bay reactor neutrino experiment


    An, F. P.; Carr, R.; McKeown, R.D.; Tsang, R. H. M.; Wu, F.F.


    The Daya Bay experiment was the first to report simultaneous measurements of reactor antineutrinos at multiple baselines leading to the discovery of ν¯e oscillations over km-baselines. Subsequent data has provided the world׳s most precise measurement of sin^2 2θ_(13) and the effective mass splitting Δm^2_(ee). The experiment is located in Daya Bay, China where the cluster of six nuclear reactors is among the world׳s most prolific sources of electron antineutrinos. Multiple antineutrino detect...

  7. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz


    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  8. Development and application of a specially designed heating system for temperature-programmed high-performance liquid chromatography using subcritical water as the mobile phase. (United States)

    Teutenberg, T; Goetze, H-J; Tuerk, J; Ploeger, J; Kiffmeyer, T K; Schmidt, K G; Kohorst, W gr; Rohe, T; Jansen, H-D; Weber, H


    A specially designed heating system for temperature-programmed HPLC was developed based on experimental measurements of eluent temperature inside a stainless steel capillary using a very thin thermocouple. The heating system can be operated at temperatures up to 225 degrees C and consists of a preheating, a column heating and a cooling unit. Fast cycle times after a temperature gradient can be realized by an internal silicone oil bath which cools down the preheating and column heating unit. Long-term thermal stability of a polybutadiene-coated zirconium dioxide column has been evaluated using a tubular oven in which the column was placed. The packing material was stable after 50h of operation at 185 degrees C. A mixture containing four steroids was separated at ambient conditions using a mobile phase of 25% acetonitrile:75% deionized water and a mobile phase of pure deionized water at 185 degrees C using the specially designed heating system and the PBD column. Analysis time could be drastically reduced from 17 min at ambient conditions and a flow rate of 1 mL/min to only 1.2 min at 185 degrees C and a flow rate of 5 mL/min. At these extreme conditions, no thermal mismatch was observed and peaks were not distorted, thus underlining the performance of the developed heating system. Temperature programming was performed by separating cytostatic and antibiotic drugs with a temperature gradient using only water as the mobile phase. In contrast to an isocratic elution of this mixture at room temperature, overall analysis time could be reduced two-fold from 20 to 10 min.

  9. Development of Operational Safety Monitoring System and Emergency Preparedness Advisory System for CANDU Reactors (I)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Ryu, Yong Ho [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Son, Han Seong; Song, Deok Yong [ENESYS, Daejeon (Korea, Republic of)


    As increase of operating nuclear power plants, an accident monitoring system is essential to ensure the operational safety of nuclear power plant. Thus, KINS has developed the Computerized Advisory System for a Radiological Emergency (CARE) system to monitor the operating status of nuclear power plant continuously. However, during the accidents or/and incidents some parameters could not be provided from the process computer of nuclear power plant to the CARE system due to limitation of To enhance the CARE system more effective for CANDU reactors, there is a need to provide complement the feature of the CARE in such a way to providing the operating parameters using to using safety analysis tool such as CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors. In this study, to enhance the safety monitoring measurement two computerized systems such as a CANDU Operational Safety Monitoring System (COSMOS) and prototype of CANDU Emergency Preparedness Advisory System (CEPAS) are developed. This study introduces the two integrated safety monitoring system using the R and D products of the national mid- and long-term R and D such as CISAS and ISSAC code.

  10. System and safety studies of accelerator driven systems and generation IV reactors for transmutation of minor actinides. Annual report 2009

    Energy Technology Data Exchange (ETDEWEB)

    Bergloef, Calle; Fokau, Andrei; Jolkkonen, Mikael; Tesinsky, Milan; Wallenius, Janne; Youpeng Zhang (Div. of Reactor Physics, Royal Institute of Technology, Stockholm (Sweden))


    During 2009, the reactor physics division has made a design study of a source efficient ADS with nitride fuel and 15/15Ti cladding, based on the EFIT design made within the EUROTRANS project. It was shown that the source efficiency may be doubled as compared to the reference design with oxide fuel and T91 cladding. Transient analysis of a medium sized sodium cooled reactor with MOX fuel allowed to define criteria in terms of power penalty for americium introduction. It was shown that for each percent of americium added to the fuel, the linear rating must be reduced by 6% in order for the fuel to survive postulated unprotected transients. The Sjoestrand area ratio method for reactivity determination has been evaluated experimentally in the strongly heterogeneous subcritical facility YALINA-Booster. Surprisingly, it has been found that the area ratio reactivity estimates may differ by a factor of two depending on detector position. It is shown that this strong spatial dependence can be explained based on a two-region point kinetics model and rectified by means of correction factors obtained through Monte Carlo simulations. For the purpose of measuring high energy neutron cross sections at the SCANDAL facility in Uppsala, Monte Carlo simulations of neutron to proton conversion efficiencies in CsI detectors have been performed. A uranium fuel fabrication laboratory has been taken into operation at KTH in 2009. Uranium and zirconium nitride powders have been fabricated by hydridation/nitridation of metallic source materials. Sample pellets have been pressed and ZrN discs have been sintered to 93% density by means of spark plasma sintering methods

  11. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    Directory of Open Access Journals (Sweden)

    Matthew Bucknor


    Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

  12. Advanced reactor passive system reliability demonstration analysis for an external event

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)


    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

  13. High strength sewage treatment in a UASB reactor and an integrated UASB-digester system. (United States)

    Mahmoud, Nidal


    The treatment of high strength sewage was investigated in a one-stage upflow anaerobic sludge blanket (UASB) reactor and a UASB-digester system. The one-stage UASB reactor was operated in Palestine at a hydraulic retention time (HRT) of 10h and at ambient air temperature for a period of more than a year in order to asses the system response to the Mediterranean climatic seasonal temperature fluctuation. Afterwards, the one-stage UASB reactor was modified to a UASB-digester system by incorporating a digester operated at 35 degrees C. The achieved removal efficiencies in the one-stage UASB reactor for total, suspended, colloidal, dissolved and VFA COD were 54, 71, 34, 23%, and -7%, respectively during the first warm six months of the year, and achieved only 32% removal efficiency for COD total over the following cold six months of the year. The modification of the one-stage UASB reactor to a UASB-digester system had remarkably improved the UASB reactor performance as the UASB-digester achieved removal efficiencies for total, suspended, colloidal, dissolved and VFA COD of 72, 74, 74, 62 and 70%. Therefore, the anaerobic treatment of high strength sewage during the hot period in Palestine in a UASB-digester system is very promising.

  14. Method and apparatus for enhancing reactor air-cooling system performance (United States)

    Hunsbedt, A.


    An enhanced decay heat removal system is disclosed for removing heat from the inert gas-filled gap space between the reactor vessel and the containment vessel of a liquid metal-cooled nuclear reactor. Multiple cooling ducts in flow communication with the inert gas-filled gap space are incorporated to provide multiple flow paths for the inert gas to circulate to heat exchangers which remove heat from the inert gas, thereby introducing natural convection flows in the inert gas. The inert gas in turn absorbs heat directly from the reactor vessel by natural convection heat transfer. 6 figs.

  15. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (United States)


    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  16. Subcritical water extraction of lipids from wet algal biomass (United States)

    Deng, Shuguang; Reddy, Harvind K.; Schaub, Tanner; Holguin, Francisco Omar


    Methods of lipid extraction from biomass, in particular wet algae, through conventionally heated subcritical water, and microwave-assisted subcritical water. In one embodiment, fatty acid methyl esters from solids in a polar phase are further extracted to increase biofuel production.

  17. Production of value added materials by subcritical water hydrolysis ...

    African Journals Online (AJOL)

    The aim of this study was the determination of the best experimental conditions for the production of useful materials such as amino acids by subcritical water hydrolysis from supercritical carbon dioxide extracted krill residues and to compare the results with raw krill. Subcritical water hydrolysis efficiency from raw and ...

  18. Subcritical water extraction of bioactive compounds from dry loquat ...

    African Journals Online (AJOL)


    concentrated in a rotary evaporator at 60°C until dry. The total extraction yield was obtained by the mean value of the total extracts divided by the mass of dry loquat leaves used. Subcritical water extraction. Subcritical water extraction was carried using an extractor. (Hangzhou Huali Co. Ltd, Hangzhou, China). The extractor ...

  19. Fossil-fuel processing technical/professional services: comparison of Fischer-Tropsch reactor systems. Phase I, final report

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, G.J.; Riekena, M.L.; Vickers, A.G.


    The Fischer-Tropsch reaction was commercialized in Germany and used to produce military fuels in fixed bed reactors. It was recognized from the start that this reactor system had severe operating and yield limitations and alternative reactor systems were sought. In 1955 the Sasol I complex, using an entrained bed (Synthol) reactor system, was started up in South Africa. Although this reactor was a definite improvement and is still operating, the literature is filled with proponents of other reactor systems, each claiming its own advantages. This report provides a summary of the results of a study to compare the development potential of three of these reactor systems with the commercially operating Synthol-entrained bed reactor system. The commercial Synthol reactor is used as a benchmark against which the development potential of the other three reactors can be compared. Most of the information on which this study is based was supplied by the M.W. Kellogg Co. No information beyond that in the literature on the operation of the Synthol reactor system was available for consideration in preparing this study, nor were any details of the changes made to the original Synthol system to overcome the operating problems reported in the literature. Because of conflicting claims and results found in the literature, it was decided to concentrate a large part of this study on a kinetic analysis of the reactor systems, in order to provide a theoretical analysis of intrinsic strengths and weaknesses of the reactors unclouded by different catalysts, operating conditions and feed compositions. The remainder of the study considers the physical attributes of the four reactor systems and compares their respective investment costs, yields, catalyst requirements and thermal efficiencies from simplified conceptual designs.

  20. A system dynamics model for tritium cycle of pulsed fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Zuolong; Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Chen, Dehong, E-mail: [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)


    As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

  1. Solvent tailoring in coal liquefaction. Quarterly report, May 1982-August 1982. [Comparison of subcritical and supercritical conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tarrer, A.R.; Guin, J.A.; Curtis, C.W.; Williams, D.C.


    The initial objective of this work was to study the phase distribution of donor solvents and solvent mixtures during the liquefaction of coal, to investigate the effects of phase distribution on coal conversion, and to determine the advantages, if any, of operating at subcritical and/or supercritical conditions. Computer simulations were used to predict the phase distribution, for various binary systems, as a function of temperature. The FLASH program was used to theoretically predict phase distribution for various model systems. Due to limitations in the computer program, success was achieved only in a few cases. Even in these cases, the existence of two-phase regions was observed only at temperatures and pressures far below normal liquefaction conditions. An extensive review of the literature was carried out in order to survey methods of experimentally studying vapor-liquid equilibria. Finally, some preliminary laboratory studies were carried out with the use of benzothiophene-dodecane as the model reaction system. It was felt that the study of the effect of reactor configuration on conversion would provide insight into whether phase distribution or mass transfer was the limiting consideration for coal conversion. However, no conclusive results were obtained from these studies.

  2. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (United States)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.


    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  3. Development of a nuclear reactor control system simulator using virtual instruments

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  4. Challenges to deployment of twenty-first century nuclear reactor systems. (United States)

    Ion, Sue


    The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors.

  5. Challenges to deployment of twenty-first century nuclear reactor systems (United States)

    Ion, Sue


    The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors.

  6. Challenges to deployment of twenty-first century nuclear reactor systems (United States)


    The science and engineering of materials have always been fundamental to the success of nuclear power to date. They are also the key to the successful deployment and operation of a new generation of nuclear reactor systems and their associated fuel cycles. This article reflects on some of the historical issues, the challenges still prevalent today and the requirement for significant ongoing materials R&D and discusses the potential role of small modular reactors. PMID:28293142

  7. Numerical simulation of Venturi ejector reactor in yellow phosphorus purification system

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiao-jing; Tang, Lei, E-mail:; Jiang, Zeng


    Highlights: • Venturi ejector reactor is used in yellow phosphorus purification system to obtain high purity phosphorus. • We study the changes of vacuum region and the performances of Venturi ejector reactor with different operating pressure. • The whole study is aim to investigate the operating conditions, rather than to find out the small details of the chemical reaction. - Abstract: A novel type of Venturi ejector reactor, which was used in a pilot plant test in a factory in Guizhou in China, was developed to overcome the insufficiency of chemical reaction in the stirred-tank reactor in yellow phosphorus purification system. The effects of different working medium, the changes of vacuum region, and the performances of the Venturi ejector reactor with different operating pressure were investigated by FLUENT. Results show that the absolute value of vacuum pressure of single-phase flow was smaller than two-phase flow at the same operating conditions, which meat two-phase flow has a higher suction capability. Reflow phenomena occurred near the exit of suction pipe and nozzle. The former reflow which leads to energy loss of vacuum region was undesirable, and the latter was beneficial to the dispersion of liquid yellow phosphorus. With a flow rate ratio below 0.45, the performance of the Venturi ejector reactor was effective. By adjusting the operating pressure, a proper flow rate ratio could be satisfied to meet the production needs in yellow phosphorus purification system.

  8. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)


    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  9. Graphic-object information system {open_quotes}research base for reactor materials science{close_quotes}

    Energy Technology Data Exchange (ETDEWEB)

    Markina, N.V.; Lebedeva, E.E.; Arkhangel`skii, N.V.; Semenov, S.B.; Moiseev, A.L.


    An information system developed for reactor materials research is described. The information system incorporates an expert system, MATREKS, and a heirarchial data base. The data base contains information from 20 Russian research reactors. The information system structure, data base structure, search methods, system output modes, and technical facilities and software required are briefly discussed. 6 refs., 2 figs.

  10. Reference reactor module for NASA's lunar surface fission power system

    Energy Technology Data Exchange (ETDEWEB)

    Poston, David I [Los Alamos National Laboratory; Kapernick, Richard J [Los Alamos National Laboratory; Dixon, David D [Los Alamos National Laboratory; Werner, James [INL; Qualls, Louis [ORNL; Radel, Ross [SNL


    Surface fission power systems on the Moon and Mars may provide the first US application of fission reactor technology in space since 1965. The Affordable Fission Surface Power System (AFSPS) study was completed by NASA/DOE to determine the cost of a modest performance, low-technical risk surface power system. The AFSPS concept is now being further developed within the Fission Surface Power (FSP) Project, which is a near-term technology program to demonstrate system-level TRL-6 by 2013. This paper describes the reference FSP reactor module concept, which is designed to provide a net power of 40 kWe for 8 years on the lunar surface; note, the system has been designed with technologies that are fully compatible with a Martian surface application. The reactor concept uses stainless-steel based. UO{sub 2}-fueled, pumped-NaK fission reactor coupled to free-piston Stirling converters. The reactor shielding approach utilizes both in-situ and launched shielding to keep the dose to astronauts much lower than the natural background radiation on the lunar surface. The ultimate goal of this work is to provide a 'workhorse' power system that NASA can utilize in near-term and future Lunar and Martian mission architectures, with the eventual capability to evolve to very high power, low mass systems, for either surface, deep space, and/or orbital missions.

  11. Analysis of space reactor system components: Investigation through simulation and non-nuclear testing (United States)

    Bragg-Sitton, Shannon M.

    The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and

  12. Plutonium Critical Mass Curve Comparison to Mass at Upper Subcritical Limit (USL) Using Whisper

    Energy Technology Data Exchange (ETDEWEB)

    Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes; Zhang, Ning [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Criticality Safety Division


    Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the MCNP® Monte Carlo radiation transport package. Standard approaches to validation rely on the selection of benchmarks based upon expert judgment. Whisper uses sensitivity/uncertainty (S/U) methods to select relevant benchmarks to a particular application or set of applications being analyzed. Using these benchmarks, Whisper computes a calculational margin. Whisper attempts to quantify the margin of subcriticality (MOS) from errors in software and uncertainties in nuclear data. The combination of the Whisper-derived calculational margin and MOS comprise the baseline upper subcritical limit (USL), to which an additional margin may be applied by the nuclear criticality safety analyst as appropriate to ensure subcriticality. A series of critical mass curves for plutonium, similar to those found in Figure 31 of LA-10860-MS, have been generated using MCNP6.1.1 and the iterative parameter study software, WORM_Solver. The baseline USL for each of the data points of the curves was then computed using Whisper 1.1. The USL was then used to determine the equivalent mass for plutonium metal-water system. ANSI/ANS-8.1 states that it is acceptable to use handbook data, such as the data directly from the LA-10860-MS, as it is already considered validated (Section 4.3 4) “Use of subcritical limit data provided in ANSI/ANS standards or accepted reference publications does not require further validation.”). This paper attempts to take a novel approach to visualize traditional critical mass curves and allows comparison with the amount of mass for which the keff is equal to the USL (calculational margin + margin of subcriticality). However, the intent is to plot the critical mass data along with USL, not to suggest that already accepted handbook data should have new and more rigorous requirements for validation.

  13. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion (United States)

    George, Jeffrey A.


    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  14. Selection of power plant elements for future reactor space electric power systems

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.; Bennett, G.A.; Copper, K.


    Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected.

  15. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.


    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs.

  16. On fusion driven systems (FDS) for transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Aagren, O (Uppsala Univ., Aangstroem laboratory, div. of electricity, Uppsala (Sweden)); Moiseenko, V.E. (Inst. of Plasma Physics, National Science Center, Kharkov Inst. of Physics and Technology, Kharkov (Ukraine)); Noack, K. (Forschungszentrum Dresden-Rossendorf (Germany))


    This report gives a brief description of ongoing activities on fusion driven systems (FDS) for transmutation of the long-lived radioactive isotopes in the spent nuclear waste from fission reactors. Driven subcritical systems appears to be the only option for efficient minor actinide burning. Driven systems offer a possibility to increase reactor safety margins. A comparatively simple fusion device could be sufficient for a fusion-fission machine, and transmutation may become the first industrial application of fusion. Some alternative schemes to create strong fusion neutron fluxes are presented


    Energy Technology Data Exchange (ETDEWEB)

    TOFFER, H.


    With the shutdown of the Hanford PUREX (Plutonium-Uranium Extraction Plant) reprocessing plant in the 1970s, adequate storage capacity for spent Hanford N Reactor fuel elements in the K and N Reactor pools became a concern. To maximize space utilization in the pools, accounting for fuel burnup was considered. Fuel that had experienced a neutron environment in a reactor is known as spent, exposed, or irradiated fuel. In contrast fuel that has not yet been placed in a reactor is known as green, unexposed, or unirradiated fuel. Calculations indicated that at typical fuel exposures for N Reactor, the spent-fuel critical mass would be twice the critical mass for green fuel. A decision was reached to test the calculational result with a definitive experiment. If the results proved positive, storage capacity could be increased and N Reactor operation could be prolonged. An experiment to be conducted in the N Reactor spent-fuel storage pool was designed and assembled (References 1 and 2) and the services of the Battelle Northwest Laboratories (BNWL) (now Pacific Northwest National Laboratory [PNNL]) critical mass laboratory were procured for the measurements (Reference 3). The experiments were performed in April 1975 in the Hanford N Reactor fuel storage pool. The fuel elements were MKIA fuel assemblies, comprised of two concentric tubes of low-enriched metallic uranium. Two separate sets of measurements were performed: one with unirradiated fuel and one with irradiated fuel. Both the unirradiated and irradiated fuel, were measured in the same geometry. The spent-fuel MKIA assemblies had an average burnup of 2865 MWd (megawatt days)/t. A constraint was imposed restricting the measurements to a subcritical limit of k{sub eff} = 0.97. Subcritical count rate data was obtained with pulsed-neutron and approach-to-critical measurements. Ten (10) configurations with green fuel and nine (9) configurations with spent fuel are described and evaluated. Of these, three (3) green fuel

  18. Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

    Energy Technology Data Exchange (ETDEWEB)

    Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment


    An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship `Mutsu`. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

  19. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  20. Biodegradation of 2,4-Dinitrotoluene and 2,6-Dinitrotoluene in a Pilot-Scale Aerobic Slurry Reactor System (United States)


    consideration ............................................................................. 7 4 3.8.1 Design basis for a hypothetical full-scale bioslurry ...treatment system ......... 74 3.8.2 Cost elements associated with bioslurry reactor systems ............................. 78 3.8.3 Comparison with...hypothetical full-scale bioslurry treatment system The treatment cost of a slurry reactor system depends mainly on three process parameters: (1) solids

  1. Update on Small Modular Reactors Dynamic System Modeling Tool: Web Application

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation (United States)


    Previous reports focused on the development of component and system models as well as end-to-end system models using Modelica and Dymola for two advanced reactor architectures: (1) Advanced Liquid Metal Reactor and (2) fluoride high-temperature reactor (FHR). The focus of this report is the release of the first beta version of the web-based application for model use and collaboration, as well as an update on the FHR model. The web-based application allows novice users to configure end-to-end system models from preconfigured choices to investigate the instrumentation and controls implications of these designs and allows for the collaborative development of individual component models that can be benchmarked against test systems for potential inclusion in the model library. A description of this application is provided along with examples of its use and a listing and discussion of all the models that currently exist in the library.

  2. Lead-cooled hybrid reactors and fuel regeneration for energy production and incineration evolution of physical parameters and induced radiotoxicity; Capacites des reacteurs hybrides au plomb pour la production d'energie et l'incineration avec multirecyclage des combustibles evolution des parametres physiques radiotoxicites induites

    Energy Technology Data Exchange (ETDEWEB)

    David, S


    The concept of accelerator driven subcritical reactors (hybrid reactors), as re-launched in the beginning of the 1990's by C. Rubbia and C.D. Bowman, allows to open new paths in the management of radioactive wastes. This work treats, first, of the study of the neutron multiplication characteristics in a subcritical reactor core and shows the fundamental differences with critical systems and the advantages that follow. This study is based on the series of measurements performed at Cadarache (Muse experiment), the first results of which are presented. The subcritical property of an hybrid reactor makes this system very flexible and allows to foresee different uses, like the energy production or the incineration of wastes. The second part of this work deals with the Monte Carlo simulation of the capacities of fast spectrum and lead-cooled hybrid systems to produce energy by using different fuel cycles (uranium and thorium), and in the same time regenerating the fissile matter and keeping the reactivity up without any external intervention. Different types of fuel multi-recycles are considered. The results allow to quantify the advantages linked with the use of the thorium cycle, in particular in terms of radiotoxicity abatement. The study of the intermediate steps necessary to develop this reactor technology with the present day fuels (plutonium from thermal reactors and enriched uranium) proposes an efficient management of the actinides produced by today's reactors which are used as auxiliary fissile materials. Finally, the incineration of actinides at the end of the cycle (shutdown scenario) is considered and allows to describe the advantage of lead-cooled hybrid systems for the abatement of the radiotoxicity of an inventory at the end of cycle. (J.S.)

  3. Subcritical-Water Extraction of Organics from Solid Matrices (United States)

    Amashukeli, Xenia; Grunthaner, Frank; Patrick, Steven; Kirby, James; Bickler, Donald; Willis, Peter; Pelletier, Christine; Bryson, Charles


    An apparatus for extracting organic compounds from soils, sands, and other solid matrix materials utilizes water at subcritical temperature and pressure as a solvent. The apparatus, called subcritical water extractor (SCWE), is a prototype of subsystems of future instrumentation systems to be used in searching for organic compounds as signs of past or present life on Mars. An aqueous solution generated by an apparatus like this one can be analyzed by any of a variety of established chromatographic or spectroscopic means to detect the dissolved organic compound( s). The apparatus can be used on Earth: indeed, in proof-of-concept experiments, SCWE was used to extract amino acids from soils of the Atacama Desert (Chile), which was chosen because the dryness and other relevant soil conditions there approximate those on Mars. The design of the apparatus is based partly on the fact that the relative permittivity (also known as the dielectric constant) of liquid water varies with temperature and pressure. At a temperature of 30 C and a pressure of 0.1 MPa, the relative permittivity of water is 79.6, due to the strong dipole-dipole electrostatic interactions between individual molecular dipoles. As the temperature increases, increasing thermal energy causes increasing disorientation of molecular dipoles, with a consequent decrease in relative permittivity. For example, water at a temperature of 325 C and pressure of 20 MPa has a relative permittivity of 17.5, which is similar to the relative permittivities of such nonpolar organic solvents as 1-butanol (17.8). In the operation of this apparatus, the temperature and pressure of water are adjusted so that the water can be used in place of commonly used organic solvents to extract compounds that have dissimilar physical and chemical properties.

  4. Cryogenic Cooling System for 5 kA, 200 μH Class HTS DC Reactor (United States)

    Park, Heecheol; Kim, Seokho; Kim, Kwangmin; Park, Minwon; Park, Taejun; Kim, A.-rong; Lee, Sangjin

    DC reactors, made by aluminum busbar, are used to stabilize the arc of an electric furnace. In the conventional arc furnace, the transport current is several tens of kilo-amperes and enormous resistive loss is generated. To reduce the resistive loss at the DC reactor, a HTS DC reactor can be considered. It can dramatically improve the electric efficiency as well as reduce the installation space. Similar with other superconducting devices, the HTS DC reactor requires current leads from a power source in room temperature to the HTS coil in cryogenic environment. The heat loss at the metal current leads can be minimized through optimization process considering the geometry and the transport current. However, the transport current of the HTS DC reactor for the arc furnace is much larger than most of HTS magnets and the enormous heat penetration through the current lead should be effectively removed to keep the temperature around 70∼77 K. Current leads are cooled down by circulation of liquid nitrogen from the cooling system with a stirling cryocooler. The operating temperature of HTS coil is 30∼40 K and circulation of gaseous helium is used to remove the heat generation at the HTS coil. Gaseous helium is transported through the cryogenic helium blower and a single stage GM cryocooler. This paper describes design and experimental results on the cooling system for current leads and the HTS coil of 5 kA, 200 μH class DC reactor as a prototype. The results are used to verify the design values of the cooling systems and it will be applied to the design of scale-up cooling system for 50 kA, 200 μH class DC reactor.

  5. TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

    Directory of Open Access Journals (Sweden)

    Yasmine Abdel-Maksoud


    Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

  6. Design and R&D Progress of China Lead-Based Reactor for ADS Research Facility

    Directory of Open Access Journals (Sweden)

    Yican Wu


    Full Text Available In 2011, the Chinese Academy of Sciences launched an engineering project to develop an accelerator-driven subcritical system (ADS for nuclear waste transmutation. The China Lead-based Reactor (CLEAR, proposed by the Institute of Nuclear Energy Safety Technology, was selected as the reference reactor for ADS development, as well as for the technology development of the Generation IV lead-cooled fast reactor. The conceptual design of CLEAR-I with 10 MW thermal power has been completed. KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant, key components, structural materials, fuel assembly, operation, and control. In order to validate and test the key components and integrated operating technology of the lead-based reactor, the lead alloy-cooled non-nuclear reactor CLEAR-S, the lead-based zero-power nuclear reactor CLEAR-0, and the lead-based virtual reactor CLEAR-V are under realization.

  7. The development of ex-core neutron flux monitoring system for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S


    Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

  8. Determination of the optimal positions for installing gamma ray detection systems at Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sayyah, A. [Department of Radiation Application, Shahid Beheshti University (Iran, Islamic Republic of); Rahmani, F., E-mail: [K.N. Toosi University of Technology, Department of Physics (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute (NSTRI) (Iran, Islamic Republic of)


    Dosimetric instruments must constantly monitor radiation dose levels in different areas of nuclear reactor. Tehran Research Reactor (TRR) has seven beam tubes for different research purposes. All the beam tubes extend from the reactor core to Beam Port Floor (BPF) of the reactor facility. During the reactor operation, the gamma rays exiting from each beam tube outlet produce a specific gamma dose rate field in the space of the BPF. To effectively monitor the gamma dose rates on the BPF, gamma ray detection systems must be installed in optimal positions. The selection of optimal positions is a compromise between two requirements. First, the installation positions must possess largest gamma dose rates and second, gamma ray detectors must not be saturated in these positions. In this study, calculations and experimental measurements have been carried out to identify the optimal positions of the gamma ray detection systems. Eight three dimensional models of the reactor core and related facilities corresponding to eight scenarios have been simulated using MCNPX Monte Carlo code to calculate the gamma dose equivalent rate field in the space of the BPF. These facilities are beam tubes, thermal column, pool, BPF space filled with air, facilities such as neutron radiography facility, neutron powder diffraction facility embedded in the beam tubes as well as biological shields inserted into the unused beam tubes. According to the analysis results of the combined gamma dose rate field, three positions on the north side and two positions on the south side of the BPF have been recognized as optimal positions for installing the gamma ray detection systems. To ensure the consistency of the simulation data, experimental measurements were conducted using TLDs (600 and 700) pairs during the reactor operation at 4.5 MW.

  9. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.


    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  10. An add-on system including a micro-reactor for an atr-ir spectrometer

    DEFF Research Database (Denmark)


    The invention relates to an add-on system for an attenuated total reflectance infrared (ATR-IR) spectrometer, the add-on system allowing for time-resolved in situ IR measurements of heterogeneous mixtures. The add-on device comprises a micro-reactor (300A) forming a sample cavity (305) when...

  11. Utilizing a Russian space nuclear reactor for a US space mission: Systems integration issues (United States)

    Reynolds, E.; Schaefer, E.; Polansky, G.; Lacy, J.; Bocharov, A.


    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a US spacecraft. One component is the Topaz 2 Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz 2 reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz 2 was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz 2, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons learned regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  12. Utilizing a Russian Space Nuclear Reactor for a United States Space Mission: Systems Integration Issues (United States)

    Reynolds, Edward; Schaefer, Edward; Polansky, Gary; Lacy, Jeff; Bocharov, Anatoly


    The Nuclear Electric Propulsion Space Test Program (NEPSTP) has developed a cooperative relationship with several institutes of the former Soviet Union to evaluate Russian space hardware on a U.S. spacecraft. One component is the Topaz II Nuclear Power System; a built and flight qualified nuclear reactor that has yet to be tested in space. The access to the Topaz II reactor provides the NEPSTP with a rare opportunity; to conduct an early flight demonstration of nuclear electric propulsion at a relatively low cost. This opportunity, however, is not without challenges. Topaz II was designed to be compatible with Russian spacecraft and launch vehicles. It was manufactured and flight qualified by Russian techniques and standards and conforms to safety requirements of the former Soviet Union, not the United States. As it is desired to make minimal modifications to the Topaz II, integrating the reactor system with a United States spacecraft and launch vehicle presents an engineering challenge. This paper documents the lessons learned regarding the integration of reactor based spacecraft and also some insight about integrating Russian hardware. It examines the planned integration flow along with specific reactor requirements that affect the spacecraft integration including American-Russian space system compatibility.

  13. Design Improvement for the Reactor Trip Switchgear System for APR1400 Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han Gyu; Choi, Woong Seock; Sohn, Se Do [KEPCO Engineering and Construction Co., Deajeon (Korea, Republic of)


    The Reactor Trip Switchgear System (RTSS) performs the function to open the Reactor Trip Circuit Breaker (RTCB) when the RTSS receives trip signals from the Plant Protection System (PPS). The RTSS for Shin-Hanul Nuclear Power Plant Units 1 and 2 (SHN 1 and 2) receives the reactor trip signals from four independent PPS divisions and performs the function to interrupt power from the Motor Generator Set (MG Set) to the Digital Rod Control System (DRCS). The RTSS for SHN 1 and 2 consists of four Reactor Trip Switchgears (RTSGs) which form the selective 2-out-of-4 logic. The RTSS design for APR 1400 DC has been changed from selective 2-out-of-4 to full 2-out-of-4 logic by configuring two independent sets of RTSS for diversity. The RTSS with the full 2-out-of-4 logic decreases the chances of generating an inadvertent reactor trip by a failure during maintenance or testing. We expect this design change to contribute to enhancing the plant availability. After all, the quantitative reliability analysis will be necessary to visualize the degree of the plant availability enhancement from the design change described in this paper.

  14. Novel, Integrated Reactor / Power Conversion System (LMR-AMTEC)

    Energy Technology Data Exchange (ETDEWEB)

    Pablo Rubiolo, Principal Investigator


    The main features of this project were the development of a long life (up to 10 years) Liquid Metal Reactor (LMR) and a static conversion subsystem comprising an Alkali Metal Thermal-to-Electric (AMTEC) topping cycle and a ThermoElectric (TE) Bottom cycle. Various coupling options of the LMR with the energy conversion subsystem were explored and, base in the performances found in this analysis, an Indirect Coupling (IC) between the LMR and the AMTEC/TE converters with Alkali Metal Boilers (AMB) was chosen as the reference design. The performance model of the fully integrated sodium-and potassium-AMTEC/TE converters shows that a combined conversion efficiency in excess of 30% could be achieved by the plant. (B204)

  15. Modeling, simulation, and analysis of a reactor system for the generation of white liquor of a pulp and paper industry

    Directory of Open Access Journals (Sweden)

    Ricardo Andreola


    Full Text Available An industrial system for the production of white liquor of a pulp and paper industry, Klabin Paraná Papéis, formed by ten reactors was modeled, simulated, and analyzed. The developed model considered possible water losses by the evaporation and reaction, in addition to variations in the volumetric flow of lime mud across the reactors due to the composition variations. The model predictions agreed well with the process measurements at the plant and the results showed that the slaking reaction was nearly complete at the third causticizing reactor, while causticizing ends by the seventh reactor. Water loss due to slaking reaction and evaporation occurred more pronouncedly in the slaker reactor than in the final causticizing reactors; nevertheless, the lime mud flow remained nearly constant across the reactors.

  16. Lunar in-core thermionic nuclear reactor power system conceptual design (United States)

    Mason, Lee S.; Schmitz, Paul C.; Gallup, Donald R.


    This paper presents a conceptual design of a lunar in-core thermionic reactor power system. The concept consists of a thermionic reactor located in a lunar excavation with surface mounted waste heat radiators. The system was integrated with a proposed lunar base concept representative of recent NASA Space Exploration Initiative studies. The reference mission is a permanently-inhabited lunar base requiring a 550 kWe, 7 year life central power station. Performance parameters and assumptions were based on the Thermionic Fuel Element (TFE) Verification Program. Five design cases were analyzed ranging from conservative to advanced. The cases were selected to provide sensitivity effects on the achievement of TFE program goals.

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

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schlagenhaufer, M.; Kohly, C.; Schulenberg, T. [Karlsruhe Inst. of Tech., Karlsruhe (Germany); Rothschmitt, S.; Bittermann, D. [AREVA NP GmbH, Erlangen (Germany)


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

  18. Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

    Energy Technology Data Exchange (ETDEWEB)

    Sponza, Delia Teresa, E-mail: [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, Tinaztepe, 35160 Izmir (Turkey); Demirden, Pinar, E-mail: [Environmental Engineer, Koza Gold Company, Environmental Department, Ovacik, Bergama Izmir (Turkey)


    In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

  19. Nuclear proliferation and civilian nuclear power: report of the Nonproliferation Alternative Systems Assessment Program. Volume IX. Reactor and fuel cycle descriptions

    Energy Technology Data Exchange (ETDEWEB)


    The Nonproliferation Alternative Systems Assessment Program (NASAP) has characterized and assessed various reactor/fuel-cycle systems. Volume IX provides, in summary form, the technical descriptions of the reactor/fuel-cycle systems studied. This includes the status of the system technology, as well as a discussion of the safety, environmental, and licensing needs from a technical perspective. This information was then used in developing the research, development, and demonstration (RD and D) program, including its cost and time frame, to advance the existing technology to the level needed for commercial use. Wherever possible, the cost data are given as ranges to reflect the uncertainties in the estimates. Volume IX is divided into three sections: Chapter 1, Reactor Systems; Chapter 2, Fuel-Cycle Systems; and the Appendixes. Chapter 1 contains the characterizations of the following 12 reactor types: light-water reactor; heavy-water reactor; water-cooled breeder reactor; high-temperature gas-cooled reactor; gas-cooled fast reactor; liquid-metal fast breeder reactor; spectral-shift-controlled reactor; accelerator-driven reactor; molten-salt reactor; gaseous-core reactor; tokamak fusion-fisson hybrid reactor; and fast mixed-spectrum reactor. Chapter 2 contains similar information developed for fuel-cycle facilities in the following categories: mining and milling; conversion and enrichment; fuel fabrication; spent fuel reprocessing; waste handling and disposal; and transportation of nuclear materials.

  20. Self locking drive system for rotating plug of a nuclear reactor (United States)

    Brubaker, James E.


    This disclosure describes a self locking drive system for rotating the plugs on the head of a nuclear reactor which is able to restrain plug motion if a seismic event should occur during reactor refueling. A servomotor is engaged via a gear train and a bull gear to the plug. Connected to the gear train is a feedback control system which allows the motor to rotate the plug to predetermined locations for refueling of the reactor. The gear train contains a self locking double enveloping worm gear set. The worm gear set is utilized for its self locking nature to prevent unwanted rotation of the plugs as the result of an earthquake. The double enveloping type is used because its unique contour spreads the load across several teeth providing added strength and allowing the use of a conventional size worm.

  1. Combustion flame-plasma hybrid reactor systems, and chemical reactant sources (United States)

    Kong, Peter C


    Combustion flame-plasma hybrid reactor systems, chemical reactant sources, and related methods are disclosed. In one embodiment, a combustion flame-plasma hybrid reactor system comprising a reaction chamber, a combustion torch positioned to direct a flame into the reaction chamber, and one or more reactant feed assemblies configured to electrically energize at least one electrically conductive solid reactant structure to form a plasma and feed each electrically conductive solid reactant structure into the plasma to form at least one product is disclosed. In an additional embodiment, a chemical reactant source for a combustion flame-plasma hybrid reactor comprising an elongated electrically conductive reactant structure consisting essentially of at least one chemical reactant is disclosed. In further embodiments, methods of forming a chemical reactant source and methods of chemically converting at least one reactant into at least one product are disclosed.

  2. Contribution of reactor physics in past and future. Is reactor physics useful?

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ. (Japan); Kosaka, Shinya [TEPCO Systems Co. (Japan); Tatsumi, Masahiro [Nuclear Fuel Industries Ltd., Tokyo (Japan)] (and others)


    Reactor Physics is a science to create rector and to play an important role in application to calculation science and safety evaluation. This feature articles contains topics, interested problems and development problems in the following field of reactor physics such as theory and experiment of reactor physics, core control, safety evaluation, criticality safety, accelerator driven subcritical reactor (ADS), new type reactor and evaluation of reactor physics. An original nuclear calculation method developed in Japan has been applied to design and analysis of fast breeder reactor. Interested problems are a proposal of fundamental principles of progressive reactor, development of calculation science, new knowledge by application of best estimate method to safety evaluation and investigation of complicated phenomena of criticality safety. (S.Y.)

  3. Design of conduction cooling system for a high current HTS DC reactor (United States)

    Dao, Van Quan; Kim, Taekue; Le Tat, Thang; Sung, Haejin; Choi, Jongho; Kim, Kwangmin; Hwang, Chul-Sang; Park, Minwon; Yu, In-Keun


    A DC reactor using a high temperature superconducting (HTS) magnet reduces the reactor’s size, weight, flux leakage, and electrical losses. An HTS magnet needs cryogenic cooling to achieve and maintain its superconducting state. There are two methods for doing this: one is pool boiling and the other is conduction cooling. The conduction cooling method is more effective than the pool boiling method in terms of smaller size and lighter weight. This paper discusses a design of conduction cooling system for a high current, high temperature superconducting DC reactor. Dimensions of the conduction cooling system parts including HTS magnets, bobbin structures, current leads, support bars, and thermal exchangers were calculated and drawn using a 3D CAD program. A finite element method model was built for determining the optimal design parameters and analyzing the thermo-mechanical characteristics. The operating current and inductance of the reactor magnet were 1,500 A, 400 mH, respectively. The thermal load of the HTS DC reactor was analyzed for determining the cooling capacity of the cryo-cooler. The study results can be effectively utilized for the design and fabrication of a commercial HTS DC reactor.

  4. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    Energy Technology Data Exchange (ETDEWEB)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey


    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  5. Astrobionibbler: In Situ Microfluidic Subcritical Water Extraction of Amino Acids (United States)

    Noell, A. C.; Fisher, A. M.; Takano, N.; Fors-Francis, K.; Sherrit, S.; Grunthaner, F.


    A fluidic-chip based instrument for subcritical water extraction (SCWE) of amino acids and other organics from powder samples has been developed. A variety of soil analog extractions have been performed to better understand SCWE capabilities.

  6. Pulsed neutron source based on accelerator-subcritical-assembly

    Energy Technology Data Exchange (ETDEWEB)

    Inoue, Makoto; Noda, Akira; Iwashita, Yoshihisa; Okamoto, Hiromi; Shirai, Toshiyuki [Kyoto Univ., Uji (Japan). Inst. for Chemical Research


    A new pulsed neutron source which consists of a 300MeV proton linac and a nuclear fuel subcritical assembly is proposed. The proton linac produces pulsed spallation neutrons, which are multipied by the subcritical assembly. A prototype proton linac that accelerates protons up to 7MeV has been developed and a high energy section of a DAW structure is studied with a power model. Halo formations in high intensity beam are also being studied. (author)

  7. On Respiratory Rate of Cherry Tomatoes under Subcritical Heights

    Directory of Open Access Journals (Sweden)

    Fang Duan


    Full Text Available The influence of subcritical drop heights on respiratory rate was studied for cherry tomatoes. The cherry tomatoes were dropped, and the mean value of O2 concentration was measured, and then the respiration rate was calculated. The results showed that the respiration rate of the cherry tomatoes increases remarkably with the dropping height. Finally, the relationship between the subcritical dropping heights and respiration rate was modeled and validated, showing good agreement.

  8. On Respiratory Rate of Cherry Tomatoes under Subcritical Heights


    Fang Duan; Yu-fen Chen; Zhong-zheng Sun; Ming-qing Chen; Hui Zhang; Jing Zhang


    The influence of subcritical drop heights on respiratory rate was studied for cherry tomatoes. The cherry tomatoes were dropped, and the mean value of O2 concentration was measured, and then the respiration rate was calculated. The results showed that the respiration rate of the cherry tomatoes increases remarkably with the dropping height. Finally, the relationship between the subcritical dropping heights and respiration rate was modeled and validated, showing good agreement.

  9. Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)


    JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

  10. Nuclear reactor with makeup water assist from residual heat removal system (United States)

    Corletti, Michael M.; Schulz, Terry L.


    A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

  11. Oxygen transport membrane reactor based method and system for generating electric power (United States)

    Kelly, Sean M.; Chakravarti, Shrikar; Li, Juan


    A carbon capture enabled system and method for generating electric power and/or fuel from methane containing sources using oxygen transport membranes by first converting the methane containing feed gas into a high pressure synthesis gas. Then, in one configuration the synthesis gas is combusted in oxy-combustion mode in oxygen transport membranes based boiler reactor operating at a pressure at least twice that of ambient pressure and the heat generated heats steam in thermally coupled steam generation tubes within the boiler reactor; the steam is expanded in steam turbine to generate power; and the carbon dioxide rich effluent leaving the boiler reactor is processed to isolate carbon. In another configuration the synthesis gas is further treated in a gas conditioning system configured for carbon capture in a pre-combustion mode using water gas shift reactors and acid gas removal units to produce hydrogen or hydrogen-rich fuel gas that fuels an integrated gas turbine and steam turbine system to generate power. The disclosed method and system can also be adapted to integrate with coal gasification systems to produce power from both coal and methane containing sources with greater than 90% carbon isolation.

  12. Recycling high-performance carbon fiber reinforced polymer composites using sub-critical and supercritical water (United States)

    Knight, Chase C.

    of the polymer matrix. To date, very few studies have been reported in this area and the studies thus far have only focused on small scale feasibility and have only shown the recovery of random fibers. The goal of this research is to advance the knowledge in the field of sub-critical and supercritical fluid recycling by providing fundamental information that will be necessary to move this process forward to an industrial scale. This dissertation work consists of several phases of studies. In the first phase of this research, the feasibility of recycling woven CFRP was established on a scale approximately 30 times larger than previously reported. The industrial relevance was also conveyed, as the process was shown to remove up 99% of a highly cross-linked resin from an aerospace grade composite system with 100% retention of the single filament tensile strength and modulus whilst also retaining the highly valuable woven fiber structure. The second phase of research demonstrated the power of this technology to recycle multi-layer composites and provide the ability to reuse the highly valuable materials. Up to 99% resin elimination was achieved for a woven 12-layer aerospace grade composite. The recycled woven fabric layers, with excellent retention of the fiber architecture, were directly reused to fabricate reclaimed fiber composites (RFC). Manufacturing issues associated with the use of the recycled fiber were investigated. Several fabrication technologies were used to fabricate the composite, and the composites show moderate short beam shear strength and may be suitable for certain industrial applications. Moreover, fresh composites were also recycled, recovered, and reused to investigate the retention of flexural properties of the fibers after recycling. Up to 95% of the flexural strength and 98% of the flexural modulus was retained in the reclaimed fiber composites. The recycled resin residual can be incorporated into fresh resin and cured, demonstrating a near

  13. Municipal waste stabilization in a reactor with an integrated active and passive aeration system. (United States)

    Kasinski, Slawomir; Slota, Monika; Markowski, Michal; Kaminska, Anna


    To test whether an integrated passive and active aeration system could be an effective solution for aerobic decomposition of municipal waste in technical conditions, a full-scale composting reactor was designed. The waste was actively aerated for 5d, passively aerated for 35 d, and then actively aerated for 5d, and the entire composting process was monitored. During the 45-day observation period, changes in the fractional, morphological and physico-chemical characteristics of the waste at the top of the reactor differed from those in the center of the reactor. The fractional and morphological analysis made during the entire process of stabilization, showed the total reduction of organic matter measured of 82 wt% and 86 wt% at the respective depths. The reduction of organic matter calculated using the results of Lost of Ignition (LOI) and Total Organic Carbon (TOC) showed, respectively, 40.51-46.62% organic matter loss at the top and 45.33-53.39% in the center of the reactor. At the end of the process, moisture content, LOI and TOC at the top were 3.29%, 6.10% and 4.13% higher, respectively, than in the center. The results showed that application of passive aeration in larger scale simultaneously allows the thermophilic levels to be maintained during municipal solid waste composting process while not inhibiting microbial activity in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  14. Sewage treatment in integrated system of UASB reactor and duckweed pond and reuse for aquaculture. (United States)

    Mohapatra, D P; Ghangrekar, M M; Mitra, A; Brar, S K


    The performance of a laboratory-scale upflow anaerobic sludge blanket (UASB) reactor and a duckweed pond containing Lemna gibba was investigated for suitability for treating effluent for use in aquaculture. While treating low-strength sewage having a chemical oxygen demand (COD) of typically less than 200 mg/L, with an increase in hydraulic retention time (HRT) from 10.04 to 33.49 h, COD removal efficiency of the UASB reactor decreased owing to a decrease in organic loading rate (OLR) causing poor mixing in the reactor. However, even at the lower OLR (0.475 kg COD/(m3 x d)), the UASB reactor gave a removal efficiency of 68% for COD and 74% for biochemical oxygen demand (BOD). The maximum COD, BOD, ammonia-nitrogen and phosphate removal efficiencies of the duckweed pond were 40.77%, 38.01%, 61.87% and 88.57%, respectively. Decreasing the OLR by increasing the HRT resulted in an increase in efficiency of the duckweed pond for removal of ammonia-nitrogen and phosphate. The OLR of 0.005 kg COD/(m2 x d) and HRT of 108 h in the duckweed pond satisfied aquaculture quality requirements. A specific growth rate of 0.23% was observed for tilapia fish fed with duckweed harvested from the duckweed pond. The economic analysis proved that it was beneficial to use the integrated system of a UASB reactor and a duckweed pond for treatment of sewage.

  15. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun


    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  16. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun


    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  17. Radiotracer investigation on the measurement of residence time distribution in an ethyl acetate reactor system with a large recycle ratio. (United States)

    Datta, Arghya; Kumar Gupta, Raj; Goswami, Sunil; Kumar Sharma, Vijay; Bhunia, Haripada; Singh, Damandeep; Jagat Pant, Harish


    A radiotracer investigation was carried out on the measurement of residence time distribution (RTD) of process fluid in an industrial-scale ethyl acetate reactor system, which consists of two independent reactors with recirculation and connected in series with each other. Bromine-82 as ammonium bromide was used as the radiotracer for the RTD experiments at different operating conditions. The individual reactors and the overall reactor system were modelled using physically representative phenomenological models comprising of continuously stirred tank reactors (CSTRs). The results showed that the recirculation rate considerably affected the flow mixing behaviour and mean residence time of the process fluid in the reactor system. The results also showed that there was bypassing of the fluid in the first reactor that ranged from 12% to 22% and 40% dead volume at different operating conditions, whereas the second reactor behaved closely as an ideal CSTR. The results of the investigation can be used to optimise the process parameters and design new improved reactor systems for the production of ethyl acetate. Copyright © 2017 Elsevier Ltd. All rights reserved.

  18. Study of DNA damage with a new system for irradiation of samples in a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gual, Maritza R., E-mail: mrgual@instec.c [Instituto Superior de Tecnologias y Ciencias Aplicadas, InSTEC, Avenida Salvador Allende y Luaces, Quinta de Los Molinos, Plaza de la Revolucion, Havana, AP 6163 (Cuba); Milian, Felix M. [Universidade Estadual de Santa Cruz, UESC (Brazil); Deppman, Airton [Instituto de Fisica, Universidad de Sao Paulo, IF-USP, Rua do Matao, Travessa R, no. 187, Ciudade Universitaria, Butanta, CEP 05508-900, Sao Paulo (Brazil); Coelho, Paulo R.P. [Instituto de Pesquisas Energeticas e Nucleares, IPEN-CNEN/SP (Brazil)


    In this paper, we report results of a quantitative analysis of the effects of neutrons on DNA, and, specifically, the production of simple and double breaks of plasmid DNA in aqueous solutions with different concentrations of free-radical scavengers. The radiation damage to DNA was evaluated by electrophoresis through agarose gels. The neutron and gamma doses were measured separately with thermoluminescent detectors. In this work, we have also demonstrated usefulness of a new system for positioning and removing samples in channel BH3 of the IEA-R1 reactor at the Instituto de Pesquisas Energeticas e Nucleares (Brazil) without necessity of interrupting the reactor operation.

  19. Design and development of a low-temperature reactor system for biorefining waste oil


    Pedersen, Hans Olav


    The background for this master’s thesis is the focus on bioenergy and biofuels at NMBU. This has, among others, resulted in a prototype of a small-scale biorefinery, which uses methanol and waste cooking oil to produce biodiesel. The purpose of this thesis is to develop a reactor system that serves as a platform for reactors to operate on and a technological alternative for a periodically on-site clean of catalysts. The purpose of the catalysts wash is to extend their lifetime, in order to ma...

  20. Automation of the radiation protection monitoring system in the RP-10 reactor; Automatizacion del sistema de monitoraje de radioproteccion en el reactor RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Anaya G, Olgger; Castillo Y, Walter; Ovalle S, Edgar [Instituto Peruano de Energia Nuclear, Lima (Peru)


    During the reactor operation, it is necessary to carry out the radiological control in the different places of the reactor, in periodic form and to take a registration of these values. For it the radioprotection official, makes every certain periods, settled down in the procedures, to verify and to carry out the registration of those values in manual form of each one of the radiation monitors. For this reason it was carried out the design and implementation of an automatic monitoring system of radioprotection in the reactor. In the development it has been considered the installation of a acquisition data system for 27 radiation gamma monitors of the type Geiger Mueller, installed inside the different places of the reactor and in the laboratories where they are manipulated radioactive material, using as hardware the FieldPoint for the possessing and digitalization of the signs which are correspondents using the communication protocol RS-232 to a PC in which has settled a program in graphic environment that has been developed using the tools of the programming software LabWindows/CVI. Then, these same signs are sent 'on line' to another PC that is in the Emergency Center of Coordination to 500 m of the reactor, by means of a system of radiofrequency communication. (author)

  1. Overview of Progress on the EU DEMO Reactor Magnet System Design

    NARCIS (Netherlands)

    Zani, L.; Bayer, C.; biancolini, M.E.; Bonifetto, R.; Nijhuis, Arend; Yagotintsev, K.


    The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In

  2. Experimental computer-controlled instrumentation system for the research reactor DR2

    DEFF Research Database (Denmark)

    Goodstein, L.P.


    An instrumentation system has been developed for one of the Danish Atomic Energy Commission's research reactors as part of an experiment on the advantages to be gained by the use of digital computers in a process plant application. Problem areas to be investigated include (a) reliability and safety...

  3. Evaluation of the dual digestion system 2: operation and performance of the pure oxygen aerobic reactor

    CSIR Research Space (South Africa)

    Messenger, JR


    Full Text Available In a comprehensive study of the performance of a full-scale (45 m3) pure oxygen autothermal thermophilic aerobic reactor of a sewage sludge dual digestion system, it was found that: Biological heat generation rate was directly proportional...

  4. Reactor-Capaсitor Device for Flexible Link Between Non-Synchronous Power Systems

    Directory of Open Access Journals (Sweden)

    Bosneaga V.


    Full Text Available In present flexible interconnections for transmission of required active power between different power systems is used, as a rule, so-called DC back-to-back link. The aim of this work is the investigation of proposed reactor-capacitor device for flexible connection of asynchronously alternating current power systems with the same nominal values of frequencies for parallel operation. The reactor-capacitor device was elaborated. The installation develops the idea of controlled reactor alternating current link, and provides reactive power balance in the unit and needed value of the output voltage module. The basic characteristics of reactor-capacitor device for controlled power transmission were investigated. Analytical expressions for device elements parameters were derived. These ensure necessary ratio of voltages modules of linked power systems and reactive power balance of the device at circular output voltage vector rotation for a given load admittance. Obtained parameters ensure constant active power flow between linked asynchronously power systems and device reactive power internal balance.

  5. New approach to control the methanogenic reactor of a two-phase anaerobic digestion system

    Energy Technology Data Exchange (ETDEWEB)

    Sachs, J. von; Meyer, U.; Rys, P.; Feitkenhauer, H. [ETH Zurich (Switzerland). Laboratorium fuer Technische Chemie


    A new control strategy for the methanogenic reactor of a two-phase anaerobic digestion system has been developed and successfully tested on the laboratory scale. The control strategy serves the purpose to detect inhibitory effects and to achieve good conversion. The concept is based on the idea that volatile fatty acids (VFA) can be measured in the influent of the methanogenic reactor by means of titration. Thus, information on the output (methane production) and input of the methanogenic reactor is available, and a (carbon) mass balance can be obtained. The control algorithm comprises a proportional/integral structure with the ratio of (a) the methane production rate measured online and (b) a maximum methane production rate expected (derived from the stoichiometry) as a control variable. The manipulated variable is the volumetric feed rate. Results are shown for an experiment with VFA (feed) concentration ramps and for experiments with sodium chloride as inhibitor. (author)

  6. Innovative inspection system for reactor pressure vessels; Innovative Pruefsysteme fuer Reaktordruckbehaelter

    Energy Technology Data Exchange (ETDEWEB)

    Mertens, K.; Trautmann, H.


    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) [Deutsch] Die vorgestellten kleinen, flexiblen und modernen Schwimmsysteme (DELPHIN-Manipulatoren und MIDAS-U-Boote) sind innovative Systeme fuer die Reduzierung der Aufwaende und Zeit zur Pruefung des Reaktordruckbehaelters und damit zur Reduktion der Revisionszeiten der Reaktoranlagen. (orig.)

  7. Ion transport membrane reactor systems and methods for producing synthesis gas (United States)

    Repasky, John Michael


    Embodiments of the present invention provide cost-effective systems and methods for producing a synthesis gas product using a steam reformer system and an ion transport membrane (ITM) reactor having multiple stages, without requiring inter-stage reactant injections. Embodiments of the present invention also provide techniques for compensating for membrane performance degradation and other changes in system operating conditions that negatively affect synthesis gas production.

  8. Treatment of fruit-juice industry wastewater in a two-stage anaerobic hybrid (AH) reactor system followed by a sequencing batch reactor (SBR). (United States)

    Tawfik, A; El-Kamah, H


    This study has been carried out to assess the performance of a combined system consisting of an anaerobic hybrid (AH) reactor followed by a sequencing batch reactor (SBR) for treatment of fruit-juice industry wastewater at a temperature of 26 degrees C. Three experimental runs were conducted in this investigation. In the first experiment, a single-stage AH reactor was operated at a hydraulic retention time (HRT) of 10.2 h and organic loading rate (OLR) of 11.8 kg COD m(-3) d(-1). The reactor achieved a removal efficiency of 42% for chemical oxygen demand (COD), 50.8% for biochemical oxygen demand (BOD5), 50.3% for volatile fatty acids (VFA) and 56.4% for total suspended solids (TSS). In the second experiment, two AH reactors connected in series achieved a higher removal efficiency for COD (67.4%), BOD5 (77%), and TSS (71.5%) at a total HRT of 20 h and an OLR of 5.9 kg COD m(-3) d(-1). For removal of the remaining portions of COD, BOD5 and TSS from the effluent of the two-stage AH system, a sequencing batch reactor (SBR) was investigated as a post-treatment unit. The reactor achieved a substantial reduction in total COD, resulting in an average effluent concentration of 50 mg L(-1) at an HRT of 11 h and OLR of 5.3 kg COD m(-3) d(-1). Almost complete removal of total BOD5 and oil and grease was achieved, i.e. 10 mg L(-1) and 1.2 mg L(-1), respectively, remained in the final effluent of the SBR.

  9. Lasers and power systems for inertial confinement fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Stark, E.E. Jr.


    After discussing the role of lasers in ICF and the candidate lasers, several important areas of technology requirements are discussed. These include the beam transport system, the pulsed power system and the gas flow system. The system requirements, state of the art, as well as needs and prospects for new technology developments are given. Other technology issues and promising developments are described briefly.

  10. Design requirements of instrumentation and control systems for next generation reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, In Soo; Lee, Byung Sun; Park, Kwang Hyun; Park, Heu Yoon; Lee, Dong Young; Kim, Jung Taek; Hwang, In Koo; Chung, Chul Hwan; Hur, Seop; Kim, Chang Hoi; Na, Nan Ju [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    In this report, the basic design requirements of Instrumentation and Control systems for next generation reactor are described, which are top-tier level, to support the advanced I and C systems. It contains the requirements in accordance with the plant reliability, the plant performance, the operator`s aid functions, the features for maintenance and testing, licensing issues for I and C systems. Advanced I and C systems are characterized such as the application of the digital and the human engineering technologies. To development of this requirements, the I and C systems for the foreign passive and the evolutionary types of reactor and the domestic conventional reators were reviewed and anlysed. At the detail design stage, these requirements will be used for top-tier requirements. To develop the detail design requirements in the future, more quantitive and qualitive analyses are need to be added. (Author) 44 refs.

  11. HEMERA: a 3D coupled core-plant system for accidental reactor transient simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bruna, G.B.; Fouquet, F.; Dubois, F. [Institut de Radioprotection et de Surete Nucleaire, 92 - Fontenay aux Roses (France); Le Pallec, J.C.; Richebois, E.; Hourcade, E.; Poinot-Salanon, C.; Royer, E. [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S), 91 - Gif sur Yvette (France)


    In the framework of their collaboration to develop a system to study reactor transients in safety-representative conditions, IRSN (Radioprotection and Nuclear Safety Institute) and Cea have launched the development of a fully coupled 3-dimensional computational chain, called HEMERA (Highly Evolutionary Methods for Extensive Reactor Analyses), based on the French SAPHYR code system, composed of APOLLO-2, CRONOS-2 and FLICA-4 codes, and the system code CATHARE. It includes cross sections generation, steady-state, depletion and transient computation capabilities in a consistent approach. Multi-level and multi-dimensional models are developed to account for neutronics, core thermal-hydraulics, fuel thermal analysis and system thermohydraulics. Currently Control Rod Ejection (RIA) and Main Steam Line Break (MSLB) accidents are investigated. The HEMERA system is presently applied to French PWR. The present paper outlines the main physical phenomena to be accounted for in such a coupled computational chain with significant time and space effects.

  12. Total acid number reduction kinetics of naphthenic acids using non-catalytic subcritical methanol (United States)

    Zafar, Faisal; Mandal, Pradip Chandra; Shaari, Ku Zilati bt Ku; Nadeem, Saad


    Naphthenic acids (NAs) are weak organic acids present in the heavy crude oil and Oil Sand Bitumens. Whereas, the NAs are the major cause of corrosion in different processing, handling and storage equipment's of the refinery. Esterification of these acids can be an interesting method to reduce the NAs content in the crude oil besides its esters are valuable commodity and can be used as added lubricant in the oils. In this study, NAs reduction kinetics were investigated in a batch type reactor with subcritical methanol, the experiments were performed at temperatures of 150-210°C and fixed methanol partial pressure of 2 MPa. Findings of this study demonstrate that the 59% of total acid number (TAN) reduction was achieved at the temperature of 210°C, methanol partial pressure of 2 MPa and reaction time of 150 min. The TAN reduction followed second order kinetics with activation energy and frequency factor of 54.15 KJ/mol and 7.6×103, respectively. These results suggest that subcritical methanol can be an effective to reduce the TAN non-catalytically.

  13. Kinetics and reaction pathways of total acid number reduction of cyclopentane carboxylic acid using subcritical methanol

    Directory of Open Access Journals (Sweden)

    Mandal Pradip C.


    Full Text Available Cyclopentane carboxylic acid (CPCA is a model compound of Naphthenic acids (NAs. This objective of this paper is to discover total acid number (TAN reduction kinetics and pathways of the reaction between CAPA and subcritical methanol (SubC-MeOH. The experiments were carried out in an autoclave reactor at temperatures of 180-220°C, a methanol partial pressure (MPP of 3 MPa, reaction times of 0-30 min and CPCA initial gas phase concentrations of 0.016-0.04 g/mL. TAN content of the samples were analyzed using ASTM D 974 techniques. The reaction products were identified and quantified with the help of GC/MS and GC-FID respectively. Experimental results reveal that TAN removal kinetics followed first order kinetics with an activation energy of 13.97 kcal/mol and a pre-exponential factor of 174.21 s-1. Subcritical methanol is able to reduce TAN of CPCA decomposing CPCA into new compounds such as cyclopentane, formaldehyde, methyl acetate and 3-pentanol.

  14. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail:, E-mail:, E-mail: [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)


    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  15. Study of reactor Brayton power systems for nuclear electric spacecraft (United States)


    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  16. Technical report on implementation of reactor internal 3D modeling and visual database system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)


    In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

  17. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100


    Zhao, Pengcheng; Shi, Kangli; Li, Shuzhou; Feng, Jingchao; Chen, Hongli


    Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kineti...

  18. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Stansbury, C.; Taylor, C. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)


    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  19. The under-critical reactors physics for the hybrid systems; La physique des reacteurs sous-critiques des systemes hybrides

    Energy Technology Data Exchange (ETDEWEB)

    Schapira, J.P. [Institut de Physique Nucleaire, IN2P3/CNRS 91 - Orsay (France); Vergnes, J. [Electricite de France, EDF, Direction des Etudes et Recherches, 75 - Paris (France); Zaetta, A. [CEA/Saclay, Direction des Reacteurs Nucleaires, DRN, 91 - Gif-sur-Yvette (France)] [and others


    This day, organized by the SFEN, took place at Paris the 12 march 1998. Nine papers were presented. They take stock on the hybrid systems and more specifically the under-critical reactors. One of the major current preoccupation of nuclear industry is the problems of the increase of radioactive wastes produced in the plants and the destruction of the present stocks. To solve these problems a solution is the utilisation of hybrid systems: the coupling of a particle acceleration to an under-critical reactor. Historical aspects, advantages and performances of such hybrid reactors are presented in general papers. More technical papers are devoted to the spallation, the MUSE and the TARC experiments. (A.L.B.)

  20. Scaleable, High Efficiency Microchannel Sabatier Reactor Project (United States)

    National Aeronautics and Space Administration — A Microchannel Sabatier Reactor System (MSRS) consisting of cross connected arrays of isothermal or graded temperature reactors is proposed. The reactor array...

  1. The Dynamic Anaerobic Reactor & Integrated Energy System (DARIES) model: model development, validation, and sensitivity analysis. (United States)

    Brouwer, A F; Grimberg, S J; Powers, S E


    The Dynamic Anaerobic Reactor & Integrated Energy System (DARIES) model has been developed as a biogas and electricity production model of a dairy farm anaerobic digester system. DARIES, which incorporates the Anaerobic Digester Model No. 1 (ADM1) and simulations of both combined heat and power (CHP) and digester heating systems, may be run in either completely mixed or plug flow reactor configurations. DARIES biogas predictions were shown to be statistically coincident with measured data from eighteen full-scale dairy operations in the northeastern United States. DARIES biogas predictions were more accurate than predictions made by the U.S. AgSTAR model FarmWare 3.4. DARIES electricity production predictions were verified against data collected by the NYSERDA DG/CHP Integrated Data System. Preliminary sensitivity analysis demonstrated that DARIES output was most sensitive to influent flow rate, chemical oxygen demand (COD), and biodegradability, and somewhat sensitive to hydraulic retention time and digester temperature.

  2. The Pressure Relief System Design for Industrial Reactors

    Directory of Open Access Journals (Sweden)

    Iztok Hace


    Full Text Available A quick and simple approach for reactor—emergency relief system design—for runaway chemical reactions is presented. A cookbook for system sizing with all main characteristic dimensions and parameters is shown on one realistic example from process industry. System design was done based on existing theories, standards, and correlations obtained from the literature, which were implemented for presented case. A simple and effective method for emergency relief system is shown, which may serve as an example for similar systems design. Obtained results may contribute to better understanding of blow down system frequently used in industrial plants, for increasing safety, decreasing explosion damage, and alleviating the ecological problems together with environmental pollution in case of industrial accidents.

  3. Multi-unit Operations in Non-Nuclear Systems: Lessons Learned for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    OHara J. M.; Higgins, J.; DAgostino, A.


    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. Small modular reactors (SMRs) are one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants and may be operated quite differently. One difference is that multiple units may be operated by a single crew (or a single operator) from one control room. The U.S. Nuclear Regulatory Commission (NRC) is examining the human factors engineering (HFE) aspects of SMRs to support licensing reviews. While we reviewed information on SMR designs to obtain information, the designs are not completed and all of the design and operational information is not yet available. Nor is there information on multi-unit operations as envisioned for SMRs available in operating experience. Thus, to gain a better understanding of multi-unit operations we sought the lesson learned from non-nuclear systems that have experience in multi-unit operations, specifically refineries, unmanned aerial vehicles and tele-intensive care units. In this paper we report the lessons learned from these systems and the implications for SMRs.

  4. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)


    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  5. 3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems (United States)

    Hançerliogulları, Aybaba; Cini, Mesut


    In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49-50:689-695, 2000; Tillack et al. in Fusion Eng Des 65:215-261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794-1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3-23, 2006).

  6. Structural assessments of plate type support system for APR1400 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen, Anh Tung; Namgung, Ihn, E-mail:


    Highlights: • This paper investigates plate-type support structure for the reactor vessel of the APR 1400. • The tall column supports of APR1400 reactor challenges in seismic and severe accident events. • A plate-type support of reactor vessel was proposed and evaluated based on ASME code. • The plate-type support was assessed to show its higher rigidity than column-type. - Abstract: This paper investigates an alternative form of support structure for the reactor vessel of the APR 1400. The current reactor vessel adopts a four-column support arrangement locating on the cold legs of the vessel. Although having been successfully designed, the tall column structure challenges in seismic events. In addition, for the mitigation of severe accident consequences, the columns inhibit ex-vessel coolant flow, hence the elimination of the support columns proposes extra safety advantages. A plate-type support was proposed and evaluated for the adequacy of meeting the structural stiffness by Finite Element Analysis (FEA) approach. ASME Boiler and Pressure Vessel Code was used to verify the design. The results, which cover thermal and static structural analysis, show stresses are within allowable limits in accordance with the design code. Even the heat conduction area is increased for the plate-type of support system, the results showed that the thermal stresses are within allowable limits. A comparison of natural frequencies and mode shapes for column support and plate-type support were presented as well which showed higher fundamental frequencies for the plate-type support system resulting in greater rigidity of the support system. From the outcome of this research, the plate-type support is proven to be an alternative to current APR column type support design.

  7. Reforming results of a novel radial reactor for a solid oxide fuel cell system with anode off-gas recirculation (United States)

    Bosch, Timo; Carré, Maxime; Heinzel, Angelika; Steffen, Michael; Lapicque, François


    A novel reactor of a natural gas (NG) fueled, 1 kW net power solid oxide fuel cell (SOFC) system with anode off-gas recirculation (AOGR) is experimentally investigated. The reactor operates as pre-reformer, is of the type radial reactor with centrifugal z-flow, has the shape of a hollow cylinder with a volume of approximately 1 L and is equipped with two different precious metal wire-mesh catalyst packages as well as with an internal electric heater. Reforming investigations of the reactor are done stand-alone but as if the reactor would operate within the total SOFC system with AOGR. For the tests presented here it is assumed that the SOFC system runs on pure CH4 instead of NG. The manuscript focuses on the various phases of reactor operation during the startup process of the SOFC system. Startup process reforming experiments cover reactor operation points at which it runs on an oxygen to carbon ratio at the reactor inlet (ϕRI) of 1.2 with air supplied, up to a ϕRI of 2.4 without air supplied. As confirmed by a Monte Carlo simulation, most of the measured outlet gas concentrations are in or close to equilibrium.

  8. Application of the BEACON-TSM system to the operation of PWR reactors; Aplicacion del sistema Beacon TSM a la operacion de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

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


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

  9. Robotic dismantlement systems at the CP-5 reactor D&D project.

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, L. S.


    The Chicago Pile 5 (CP-5) Research Reactor Facility is currently undergoing decontamination and decommissioning (D&D) at the Argonne National Laboratory (ANL) Illinois site. CP-5 was the principle nuclear reactor used to produce neutrons for scientific research at Argonne from 1954 to 1979. The CP-5 reactor was a heavy-water cooled and moderated, enriched uranium-fueled reactor with a graphite reflector. The CP-5 D&D project includes the disassembly, segmentation and removal of all the radioactive components, equipment and structures associated with the CP-5 facility. The Department of Energy's Robotics Technology Development Program and the Federal Energy Technology Center, Morgantown Office provided teleoperated, remote systems for use in the dismantlement of the CP-5 reactor assembly for tasks requiring remote dismantlement as part of the EM-50 Large-Scale Demonstration Program (LSDP). The teleoperated systems provided were the Dual Arm Work Platform (DAWP), the Rosie Mobile Teleoperated Robot Work System (ROSIE), and a remotely-operated crane control system with installed swing-reduction control system. Another remotely operated apparatus, a Brokk BM250, was loaned to ANL by the Princeton Plasma Physics Laboratory (PPPL). This machine is not teleoperated and was not part of the LSDP, but deserves some mention in this discussion. The DAWP is a robotic dismantlement system that includes a pair of Schilling Robotic Systems Titan III hydraulic manipulator arms mounted to a specially designed support platform: a hydraulic power unit (HPU) and a remote operator console. The DAWP is designed to be crane-suspended for remote positioning. ROSIE, developed by RedZone Robotics, Inc. is a mobile, electro-hydraulic, omnidirectional platform with a heavy-duty telescoping boom mounted to the platform's deck. The work system includes the mobile platform (locomotor), a power distribution unit (PDU) and a remote operator console. ROSIE moves about the reactor building

  10. Combined on-board hydride slurry storage and reactor system and process for hydrogen-powered vehicles and devices (United States)

    Brooks, Kriston P; Holladay, Jamelyn D; Simmons, Kevin L; Herling, Darrell R


    An on-board hydride storage system and process are described. The system includes a slurry storage system that includes a slurry reactor and a variable concentration slurry. In one preferred configuration, the storage system stores a slurry containing a hydride storage material in a carrier fluid at a first concentration of hydride solids. The slurry reactor receives the slurry containing a second concentration of the hydride storage material and releases hydrogen as a fuel to hydrogen-power devices and vehicles.

  11. Subcritical water as reaction environment: fundamentals of hydrothermal biomass transformation. (United States)

    Möller, Maria; Nilges, Peter; Harnisch, Falk; Schröder, Uwe


    Subcritical water, that is, water above the boiling and below critical point, is a unique and sustainable reaction medium. Based on its solvent properties, in combination with the often considerable intrinsic water content of natural biomass, it is often considered as a potential solvent for biomass processing. Current knowledge on biomass transformation in subcritical water is, however, still rather scattered without providing a consistent picture. Concentrating on fundamental physical and chemical aspects, this review summarizes the current state of knowledge of hydrothermal biomass conversion in subcritical water. After briefly introducing subcritical water as a reaction medium, its advantages for biomass processing compared to other thermal processes are highlighted. Subsequently, the physical-chemical properties of subcritical water are discussed in the light of their impact on the occurring chemical reactions. The influence of major operational parameters, including temperature, pressure, and reactant concentration on hydrothermal biomass transformation processes are illustrated for selected carbohydrates. Major emphasis is put on the nature of the carbohydrate monomers, since the conversion of the respective polymers is analogous with the additional prior step of hydrolytic depolymerization. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  12. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo


    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  13. Development of fiber-delivered laser peening system to prevent stress corrosion cracking of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Sano, Y.; Kimura, M.; Yoda, M.; Mukai, N.; Sato, K.; Uehara, T.; Ito, T.; Shimamura, M.; Sudo, A.; Suezono, N. [Toshiba Corp., Yokohama (Japan)


    The authors have developed a system to deliver water-penetrable intense laser pulses of frequency-doubled Nd:YAG laser through optical fiber. The system is capable of improving a residual stress on water immersed metal material remotely, which is effective to prevent the initiation of stress corrosion cracking (SCC) of reactor components. Experimental results showed that a compressive residual stress with enough amplitude and depth was built in the surface layer of type 304 stainless steel (SUS304) by irradiating laser pulses through optical fiber with diameter of 1 mm. A prototype peening head with miniaturized dimensions of 88 mm x 46 mm x 25 mm was assembled to con-firm the accessibility to the heat affected zone (HAZ) along weld lines of a reactor core shroud. The accessibility was significantly improved owing to the flexible optical fiber and the miniaturized peening head. The fiber delivered system opens up the possibility of new applications of laser peening. (author)

  14. Application of fault detection and identification (FDI) techniques in power regulating systems of nuclear reactors (United States)

    Roy, K.; Banavar, R. N.; Thangasamy, S.


    Application of failure detection and identification (FDI) algorithms have essentially been limited to identification of a global fault in the system, and no further attempts have been made to locate subcomponent faults for root cause analysis. This paper presents Kalman filter-based methods for FDI in power regulating systems of nuclear reactors. The attempt here is to explain how the behavior of the states, residues, and covariances can be interpreted to identify subcomponent faults. An alternative to the Kalman filter-the risk-sensitive filter-is also introduced. Comparison of its performance with the Kalman filter-based FDI algorithms is studied. All simulation studies have been carried out on postulated faults in the power regulating system of heavy water moderated, low pressure vertical tank-type research reactors.

  15. A fancy eco-compatible wastewater treatment system: Green Bio-sorption Reactor. (United States)

    Zhao, Yaqian; Liu, Ranbin; Zhao, Jinhui; Xu, Lei; Sibille, Caroline


    A novel concept was proposed and preliminarily investigated by embedding alum sludge-based constructed wetland into conventional activated sludge system in terms of Green Bio-sorption Reactor (GBR). This novel GBR inherited the aesthetic value of constructed wetland and owned the robust phosphorus (P) adsorption along with the benefit of carriers' addition (dewatered alum sludge). The preliminary demonstration was conducted in a lab-scale sequencing batch reactor (SBR) system without biological phosphorus removal process. The novel process achieved averagely 96%, 99% and 90% for BOD, TP and TN removal with piggery wastewater as influent, demonstrating for the first time of its promising performance. Moreover, the coexistence of biofilm and suspended sludge also achieved 55-88% simultaneous nitrification and denitrification efficiency, higher than biofilm only. Overall, alum sludge-based GBR could achieve reliable pollutants removal and provides a novel and sustainable pathway to upgrade conventional activated sludge system. Copyright © 2017 Elsevier Ltd. All rights reserved.

  16. Startup thaw concept for the SP-100 space reactor power system (United States)

    Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.


    A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.


    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Laboratory; Zhang, Hongbin [Idaho National Laboratory; Zou, Ling [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory


    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  18. Model Reduction Using Proper Orthogonal Decomposition and Predictive Control of Distributed Reactor System

    Directory of Open Access Journals (Sweden)

    Alejandro Marquez


    Full Text Available This paper studies the application of proper orthogonal decomposition (POD to reduce the order of distributed reactor models with axial and radial diffusion and the implementation of model predictive control (MPC based on discrete-time linear time invariant (LTI reduced-order models. In this paper, the control objective is to keep the operation of the reactor at a desired operating condition in spite of the disturbances in the feed flow. This operating condition is determined by means of an optimization algorithm that provides the optimal temperature and concentration profiles for the system. Around these optimal profiles, the nonlinear partial differential equations (PDEs, that model the reactor are linearized, and afterwards the linear PDEs are discretized in space giving as a result a high-order linear model. POD and Galerkin projection are used to derive the low-order linear model that captures the dominant dynamics of the PDEs, which are subsequently used for controller design. An MPC formulation is constructed on the basis of the low-order linear model. The proposed approach is tested through simulation, and it is shown that the results are good with regard to keep the operation of the reactor.

  19. Pellet bed reactor concept for nuclear electric propulsion (United States)

    El-Genk, Mohamed S.; Morley, Nicholas J.; Juhasz, Albert


    For Nuclear Electric Propulsion (NEP) applications, gas cooled nuclear reactors with dynamic energy conversion systems offer high specific power and low total mass. This paper describes the Pellet Bed Reactor (PeBR) concept for potential NEP missions to Mars. The helium cooled, 75-80 MWt PeBR, consists of a single annular fuel region filled with a randomly packed bed of spherical fuel pellets, is designed for multiple starts, and offers unique safety and operation features. Each fuel pellet, about 8-10 mm in diameter, is composed of hundreds of TRISO type fuel microspheres embedded in a graphite matrix for a full retention of fission products. To eliminate the likelihood of a single-point failure, the annular core of the PeBR is divided into three 120° sectors. Each sector is self contained and separate and capable of operating and being cooled on its own and in cooperation with either one or two other sectors. Each sector is coupled to a separate, 5 MWe Closed Brayton Cycle (CBC) energy conversion unit and is subcritical for safe handling and launching. In the event of a failure of the cooling system of a core sector, the reactor power level may be reduced, allowing adjacent sectors to convect the heat away using their own cooling system, thus maintaining reactor operation. Also, due to the absence of an internal core structure in the PeBR core, fueling of the reactor can easily be performed either at the launch facility or in orbit, and refueling can be accomplished in orbit as needed to extend the power system lifetime

  20. Advances in high rate anaerobic treatment: staging of reactor systems.

    NARCIS (Netherlands)

    Lier, van J.B.; Zee, van der F.P.; Tan, N.C.G.; Rebac, S.; Kleerebezem, R.


    Anaerobic wastewater treatment (AnWT) is considered as the most cost-effective solution for organically polluted industrial waste streams. Particularly the development of high-rate systems, in which hydraulic retention times are uncoupled from solids retention times, has led to a world-wide

  1. Reactivity Monitoring of Accelerator-Driven Nuclear Reactor Systems

    NARCIS (Netherlands)

    Uyttenhove, W.


    This thesis provides a methodology and set-up of a reactivity monitoring tool for Accelerator-Driven Systems (ADS). The reactivity monitoring tool should guarantee the operation of an ADS at a safe margin from criticality. Robustness is assured in different aspects of the monitoring tool: the choice

  2. Development of Environmentally-Assisted Fatigue Monitoring System for Advanced Power Reactors (APR1400)

    Energy Technology Data Exchange (ETDEWEB)

    Park, June Soo; Kim, Yeon Jeong; Kang, Sun Yeh; Yoon, Ki Seok; Choi, Taek Sang [KEPCO-E and C, Daejeon (Korea, Republic of)


    This paper introduces an EAF monitoring system developed for Shin-Kori Nuclear Power Plant (NPP), Units 3 and 4 which are the first two reactors of the APR1400 model. The EAF monitoring system has been developed for Shin-Kori NPP, Units 3 and 4, and is ready for an application for the plant lifetime. It is expected that the plant fatigue management can be effectively fulfilled, and the structural integrity of the critical components assured by an implementation of the fatigue monitoring system from the beginning of the lifetime. When fatigue analyses including the effects of the Light-Water Reactor (LWR) environment are applicable, plant designers address the environmentally-assisted fatigue (EAF) for Class 1 reactor pressure boundary components. The environment factor (F{sub en}) method has been endorsed by the U. S. Nuclear Regulatory Commission for evaluating fatigue analyses to address the environmental effects, and this method considers four major variables in addition to the traditional air-fatigue analyses: Material temperature, dissolved oxygen content of coolant, sulfur (S) content of material, and strain rate at the material points of interest. APR1400 nuclear power plants are designed to the requirements of the enhanced plant safety, availability and performance criteria for a 60 year design life. To better manage the material degradation and structural integrity of the pressure boundary components, a fatigue monitoring system has been developed for APR1400 NPPs, which is capable to monitor the EAF damage during the plant lifetime.

  3. RSYST: an integrated modular system with a data basis, for automated calculation of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ruehle, Roland [Univ. of Stuttgart (Germany)


    The integrated modular system RSYST was developed to offer the engineer and physicist a simpler method for carrying out the layout calculations for nuclear reactors. The system consists of a data basis, a control section, a data basis monitoring system, as well as an unlimited number of modules. The data transfer between individual modules is done through the data basis monitoring program by way of the central data basis. Control words of the input permit the user to control the process of any desired modules. Each module can have flexible data input to it from the data basis. By use of special modules, logical branches and loops can be carried out. The system was implemented on a CDC 6600 and partly on an IBM 360/75. At this moment, it includes 45,000 FORTRAN statements and 120 control words. Project calculations have been successfully carried out with the aid of RSYST for over three years. At this time, in addition to the general modules there exists primarily modules for reactor statistics calculations, burn-up calculations, and shielding calculations, and for the production of group constants. A start has been made to include problems of heat conduction, thermal hydraulics, reactor safety, control technology, and loop dynamics. (auth)

  4. Hydrothermal Processing of Macroalgal Feedstocks in Continuous-Flow Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Elliott, Douglas C.; Hart, Todd R.; Neuenschwander, Gary G.; Rotness, Leslie J.; Roesijadi, Guri; Zacher, Alan H.; Magnuson, Jon K.


    Wet macroalgal slurries have been converted into a biocrude by hydrothermal liquefaction (HTL) in a bench-scale continuous-flow reactor system. Carbon conversion to a gravity-separable oil product of 58.8% was accomplished at relatively low temperature (350 °C) in a pressurized (subcritical liquid water) environment (20 MPa) when using feedstock slurries with a 21.7% concentration of dry solids. As opposed to earlier work in batch reactors reported by others, direct oil recovery was achieved without the use of a solvent, and biomass trace mineral components were removed by processing steps so that they did not cause processing difficulties. In addition, catalytic hydrothermal gasification (CHG) was effectively applied for HTL byproduct water cleanup and fuel gas production from water-soluble organics. Conversion of 99.2% of the carbon left in the aqueous phase was demonstrated. Finally, as a result, high conversion of macroalgae to liquid and gas fuel products was found with low levels of residual organic contamination in byproduct water. Both process steps were accomplished in continuous-flow reactor systems such that design data for process scale-up was generated.

  5. Pressure suppression containment system for boiling water reactor (United States)

    Gluntz, Douglas M.; Nesbitt, Loyd B.


    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  6. An advanced extruder-feeder biomass liquefaction reactor system (United States)

    White, Don H.; Wolf, D.; Davenport, G.; Mathews, S.; Porter, M.; Zhao, Y.


    A unique method of pumping concentrated, viscous biomass slurries that are characteristic of biomass direct liquefaction systems was developed. A modified single-screw extruder was shown to be capable of pumping solid slurries as high as 60 weight percent wood flour in wood oil derived vacuum bottoms, as compared to only 10 to 20 weight percent wood flour in wood oil in conventional systems. During the period August, 1985 to April, 1987, a total of 18 experimental continuous biomass liquefaction runs were made using white birch feedstock. Good operability with feed rates up to 30 lb/hr covering a range of carbon monoxide, sodium carbonate catalyst, pressures from 800 to 3000 psi and temperatures from 350 C to 430 C was achieved. Crude wood oils containing 6 to 10 weight percent residual oxygen were obtained. Other wood oil characteristics are reported.

  7. Progress in hardware development for the SAFE heatpipe reactor system (United States)

    Ring, P. J.; Sayre, E. D.; van Dyke, Melissa; Houts, Mike


    Advanced Methods & Materials Company (AMM) previously fabricated the stainless steel modules for the SAFE 30 system. These earlier modules consisting of five fuel pins surrounding a heat pipe, were brazed together using a tricusp insert in the gaps between tubes to ensure maximum braze coverage. It was decided that if possible the next generations of modules, both stainless steel and refractory alloy, would be diffusion bonded together using a Hot Issostatic Pressing (HIP) process. This process was very successfully used in producing the bonded rhenium Nb-lZr fuel cladding and the heat exchanger for the SP-100 Nuclear Space System Ref. 1 & 2. In addition AMM have since refined the technology enabling them to produce very high temperature rocket thrust chambers. Despite this background the complex geometry required for the SAFE module was quite challenging. It was necessary to develop a method which could be applied for both stainless steel and refractory alloy systems. In addition the interstices between tubes had to be completely filled with the tricusp insert to avoid causing distortion of the tube shape during HIPing and provide thermal conductivity from the fuel tubes to the heat pipes. Nevertheless it was considered worth the effort since Hot Isostatic Pressing, if successful, will produce an assembly with the heat pipe completely embedded within the module such that the diffusion bonded assembly has the thermal conduction and strength equivalent to a solid structure. .

  8. System for Coupling an IEC Reactor to Ion Thrusters (United States)

    Webber, Jason; Burton, Rodney; Momoto, Hiromu; Miley, George; Richardson, Nathan


    A conceptual design for an electric-thruster-driven space ship using a D-He3 fueled Inertial Electrostatic Confinement (IEC) fusion power unit was recently developed [1]. This propulsion system uses a bank of modified NSTAR-type krypton ion thrusters (specific impulse of 16,000 sec.) giving a total thrust of 1020 N. The thrust time for a typical outer planet mission ( e.g. Jupiter) with a delta-V of 50,000 m/s is then 200 days. A key component of this concept is a traveling wave direct energy converter that converts the kinetic energy of 14-MeV fusion reaction product protons to high voltage (about 1 MV) DC electrical output. A unique step-down transformer and rectifier system condition this output for use in the ion thrusters. Details of these components, the NSTAR-thruster modifications plus a magnetic hexa-pole collimator designed to guide the emitted protons into the traveling wave converter will be described. This advanced electric thruster design offers a very high power-to-weight ratio system that is crucial for deep space propulsion. [1] George H. Miley, Hiromu Momota, R. Burton, N.Richardson, M. Coventry, and Y. Shaban, IEC Based D-He3 Fusion for Space Propulsion, Trans Am. Nuclear Society, Annual Meeting, Hollywood, FL, June 2002.

  9. Reviewing real-time performance of nuclear reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Preckshot, G.G. [Lawrence Livermore National Lab., CA (United States)


    The purpose of this paper is to recommend regulatory guidance for reviewers examining real-time performance of computer-based safety systems used in nuclear power plants. Three areas of guidance are covered in this report. The first area covers how to determine if, when, and what prototypes should be required of developers to make a convincing demonstration that specific problems have been solved or that performance goals have been met. The second area has recommendations for timing analyses that will prove that the real-time system will meet its safety-imposed deadlines. The third area has description of means for assessing expected or actual real-time performance before, during, and after development is completed. To ensure that the delivered real-time software product meets performance goals, the paper recommends certain types of code-execution and communications scheduling. Technical background is provided in the appendix on methods of timing analysis, scheduling real-time computations, prototyping, real-time software development approaches, modeling and measurement, and real-time operating systems.

  10. A bifurcation analysis of boiling water reactor on large domain of parametric spaces (United States)

    Pandey, Vikas; Singh, Suneet


    The boiling water reactors (BWRs) are inherently nonlinear physical system, as any other physical system. The reactivity feedback, which is caused by both moderator density and temperature, allows several effects reflecting the nonlinear behavior of the system. Stability analyses of BWR is done with a simplified, reduced order model, which couples point reactor kinetics with thermal hydraulics of the reactor core. The linear stability analysis of the BWR for steady states shows that at a critical value of bifurcation parameter (i.e. feedback gain), Hopf bifurcation occurs. These stable and unstable domains of parametric spaces cannot be predicted by linear stability analysis because the stability of system does not include only stability of the steady states. The stability of other dynamics of the system such as limit cycles must be included in study of stability. The nonlinear stability analysis (i.e. bifurcation analysis) becomes an indispensable component of stability analysis in this scenario. Hopf bifurcation, which occur with one free parameter, is studied here and it formulates birth of limit cycles. The excitation of these limit cycles makes the system bistable in the case of subcritical bifurcation whereas stable limit cycles continues in an unstable region for supercritical bifurcation. The distinction between subcritical and supercritical Hopf is done by two parameter analysis (i.e. codimension-2 bifurcation). In this scenario, Generalized Hopf bifurcation (GH) takes place, which separates sub and supercritical Hopf bifurcation. The various types of bifurcation such as limit point bifurcation of limit cycle (LPC), period doubling bifurcation of limit cycles (PD) and Neimark-Sacker bifurcation of limit cycles (NS) have been identified with the Floquet multipliers. The LPC manifests itself as the region of bistability whereas chaotic region exist because of cascading of PD. This region of bistability and chaotic solutions are drawn on the various

  11. Applying and adapting the Swedish regulatory system for decommissioning to nuclear power reactors - The regulator's perspective. (United States)

    Amft, Martin; Leisvik, Mathias; Carroll, Simon


    Half of the original 13 Swedish nuclear power reactors will be shut down by 2020. The decommissioning of these reactors is a challenge for all parties involved, including the licensees, the waste management system, the financing system, and the Swedish Radiation Safety Authority (SSM). This paper presents an overview of the Swedish regulations for decommissioning of nuclear facilities. It describes some of the experiences that SSM has gained from the application of these regulations. The focus of the present paper is on administrative aspects of decommissioning, such as SSM's guidelines, the definition of fundamental concepts in the regulatory framework, and a proposed revision of the licensing process according to the Environmental Act. These improvements will help to streamline the administration of the commercial nuclear power plant decommissioning projects that are anticipated to commence in Sweden in the near future. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    Directory of Open Access Journals (Sweden)

    A. M. Miri


    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  13. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)



    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  14. Priorities in the development of nuclear constants support system for reactor and shielding calculations

    Directory of Open Access Journals (Sweden)

    G.N. Manturov


    Development of the integral unified nuclear data support system and its implementation in the calculation codes will ensure not only the unification of the procedure for nuclear data preparation, which will allow enhancing reliability of their verification, but, as well, will enhance accuracy and reliability of calculation prediction of all the most important characteristics of the reactors under design, will ensure their licensing compliance, competitiveness and independence from foreign products.

  15. On the implementation of new technology modules for fusion reactor systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Franza, F., E-mail: [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Boccaccini, L.V.; Fisher, U. [Institute of Neutron Physics and Reactor Technology, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany); Gade, P.V.; Heller, R. [Institute for Technical Physics, Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen, 76344 (Germany)


    Highlights: • At KIT a new technology modules for systems code are under development. • A new algorithm for the definition of the main reactor's components is defined. • A new blanket model based on 1D neutronics analysis is described. • A new TF coil stress model based on 3D electromagnetic analysis is described. • The models were successfully benchmarked against more detailed models. - Abstract: In the frame of the pre-conceptual design of the next generation fusion power plant (DEMO), systems codes are being used from nearly 20 years. In such computational tools the main reactor components (e.g. plasma, blanket, magnets, etc.) are integrated in a unique computational algorithm and simulated by means of rather simplified mathematical models (e.g. steady state and zero dimensional models). The systems code tries to identify the main design parameters (e.g. major radius, net electrical power, toroidal field) and to make the reactor's requirements and constraints to be simultaneously accomplished. In fusion applications, requirements and constraints can be either of physics or technology kind. Concerning the latest category, at Karlsruhe Institute of Technology a new modelling activity has been recently launched aiming to develop improved models focusing on the main technology areas, such as neutronics, thermal-hydraulics, electromagnetics, structural mechanics, fuel cycle and vacuum systems. These activities started by developing: (1) a geometry model for the definition of poloidal profiles for the main reactors components, (2) a blanket model based on neutronics analyses and (3) a toroidal field coil model based on electromagnetic analysis, firstly focusing on the stresses calculations. The objective of this paper is therefore to give a short outline of these models.

  16. Mars, the Moon, and the Ends of the Earth: Autonomy for Small Reactor Power Systems

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Richard Thomas [ORNL


    In recent years, the National Aeronautics and Space Administration (NASA) has been considering deep space missions that utilize a small-reactor power system (SRPS) to provide energy for propulsion and spacecraft power. Additionally, application of SRPS modules as a planetary power source is being investigated to enable a continuous human presence for nonpolar lunar sites and on Mars. A SRPS can supply high-sustained power for space and surface applications that is both reliable and mass efficient. The use of small nuclear reactors for deep space or planetary missions presents some unique challenges regarding the operations and control of the power system. Current-generation terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a SRPS employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. While surface power systems for planetary outposts face less extreme delays and periods of isolation and may benefit from limited maintenance capabilities, considerations such as human safety, resource limitations and usage priorities, and economics favor minimizing direct, continuous human interaction with the SRPS for online, dedicated power system management. Thus, a SRPS control system for space or planetary missions must provide capabilities for operational autonomy. For terrestrial reactors, large-scale power plants remain the preferred near-term option for nuclear power generation. However, the desire to reduce reliance on carbon-emitting power sources in developing countries may lead to increased consideration of SRPS modules for local power generation in remote regions that are characterized by emerging, less established infrastructures

  17. Design issues on using FPGA-based I and C systems in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Santos, Isaac Jose A.L. dos; Lacerda, Fabio de, E-mail:, E-mail:, E-mail:, E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Div. de Engenharia Nuclear


    The FPGA (field programmable gate array) is widely used in various fields of industry. FPGAs can be used to perform functions that are safety critical and require high reliability, like in automobiles, aircraft control and assistance and mission-critical applications in the aerospace industry. With these merits, FPGAs are receiving increased attention worldwide for application in nuclear plant instrumentation and control (I and C) systems, mainly for Reactor Protection System (RPS). Reasons for this include the fact that conventional analog electronics technologies are become obsolete. I and C systems of new Reactors have been designed to adopt the digital equipment such as PLC (Programmable Logic Controller) and DCS (Distributed Control System). But microprocessors-based systems may not be simply qualified because of its complex characteristics. For example, microprocessor cores execute one instruction at a time, and an operating system is needed to manage the execution of programs. In turn, FPGAs can run without an operating system and the design architecture is inherently parallel. In this paper we aim to assess these and other advantages, and the limitations, on FPGA-based solutions, considering the design guidelines and regulations on the use of FPGAs in Nuclear Plant I and C Systems. We will also examine some circuit design techniques in FPGA to help mitigate failures and provide redundancy. The objective is to show how FPGA-based systems can provide cost-effective options for I and C systems in modernization projects and to the RMB (Brazilian Multipurpose Reactor), ensuring safe and reliable operation, meeting licensing requirements, such as separation, redundancy and diversity. (author)

  18. Convective wave front locking for a reaction-diffusion system in a conical flow reactor

    DEFF Research Database (Denmark)

    Kuptsov, P.V.; Kuznetsov, S.P.; Knudsen, Carsten


    We consider reaction-diffusion instabilities in a flow reactor whose cross-section slowly expands with increasing longitudinal coordinate (cone shaped reactor). Due to deceleration of the flow in this reactor, the instability is convective near the inlet to the reactor and absolute at the downstr......We consider reaction-diffusion instabilities in a flow reactor whose cross-section slowly expands with increasing longitudinal coordinate (cone shaped reactor). Due to deceleration of the flow in this reactor, the instability is convective near the inlet to the reactor and absolute...

  19. Modeling of adsorber/desorber/catalytic reactor system for ethylene oxide removal

    Directory of Open Access Journals (Sweden)



    Full Text Available The removal of ethylene oxide (EtO in a combined system adsorber/desorber/catalytic reactor has been investigated. The combined system was a modified draft tube spouted bed reactor loaded with Pt/Al2O3 catalyst. The annular region was divided into two sectons, the “hot” section contained about 7 % of catalyst and it behaved as a desorber and catalytic incinerator, while the “cold” section, with the rest of the catalyst, behaved as a sorber. The catalyst particles were circulated between the two sections by use of a draft tube riser. The Computational Fluid Dynamics (CFD program package FLUENT was used for simulations of the operation of the combined system. In addition, a one-dimensional numerical model for the operation of the packed bed reactor was compared with the corresponding FLUENT calculations. The results of the FLUENT simulations are in very good agreement with the experimental observations, as well as with the results of the one-dimensional numerical simulations.

  20. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.


    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  1. A survey of commercially available manipulators, end-effectors, and delivery systems for reactor decommissioning activities

    Energy Technology Data Exchange (ETDEWEB)

    Henley, D.R. [Argonne National Lab., IL (United States); Litka, T.J. [Advanced Consulting Group, Chicago, IL (United States)


    Numerous nuclear facilities owned by the U.S. Department of Energy (DOE) are under consideration for decommissioning. Currently, there are no standardized, automated, remote systems designed to dismantle and thereby reduce the size of activated reactor components and vessels so that they can be packaged and shipped to disposal sites. Existing dismantling systems usually consist of customized, facility-specific tooling that has been developed to dismantle a specific reactor system. Such systems have a number of drawbacks. Generally, current systems cannot be disassembled, moved, and reused. Developing and deploying the tooling for current systems is expensive and time-consuming. In addition, the amount of manual work is significant because long-handled tools must be used; as a result, personnel are exposed to excessive radiation. A standardized, automated, remote system is therefore needed to deliver the tooling necessary to dismantle nuclear facilities at different locations. Because this system would be reusable, it would produce less waste. The system would also save money because of its universal design, and it would be more reliable than current systems.

  2. Reactivity analysis for numerical solution of the point kinetic equation for subcritical; Analise da reatividade para solucao numerica da equacao da cinetica pontual para sistemas subcriticos

    Energy Technology Data Exchange (ETDEWEB)

    Henrice Junior, Edson; Goncalves, Alessandro da Cruz, E-mail:, E-mail: [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Energia Nuclear; Palma, Daniel Artur Pinheiro, E-mail: [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Mesquita, Amir Zacarias, E-mail: [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)


    This paper provides a comparison between the reactivity calculated by the approximation based on the multiplication factor (K{sub eff}) and a new approach for the reactivity calculation to be used in the kinetics point equation for subcritical systems. To obtain the necessary kinetic parameters as well and the reference Power value calculation and validation, a subcritical system was simulated with the Monte Carlo code Serpent. This study is important for determining nuclear Power in such systems. The results shown consistent values with the validation method and new in-depth studies to calculate the reactivity should be performed to such systems, making the issue a very current theme. (author)

  3. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail:, E-mail:, E-mail: [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)


    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  4. Advanced Fusion Reactors for Space Propulsion and Power Systems (United States)

    Chapman, John J.


    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  5. Assessment of the reliability of neutronic parameters of Ghana Research Reactor-1 control systems

    Energy Technology Data Exchange (ETDEWEB)

    Amponsah-Abu, E.O., E-mail: [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana); Gbadago, J.K. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana); Akaho, E.H.K.; Akoto-Bamford, S. [School of Nuclear and Allied Sciences, University of Ghana (Ghana); Gyamfi, K.; Asamoah, M.; Baidoo, I.K. [National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG. 80, Legon-Accra (Ghana)


    Highlights: • The reliability of neutronics parameters of GHARR-I was assessed. • The reactor was operated at different power levels of 5–30 kW. • The pre-set flux was compared with the flux in the inner irradiation site. • Decrease in the core reactivity caused difference in flux on the meters and site. • Neutronic parameters become reliable when operation is done at reactivity of 4 mk. - Abstract: The Ghana Research Reactor-1 (GHARR-1) has been in operation for the past 19 years using a Micro-Computer Closed Loop System (MCCLS) and Control Console (CC) as the control systems. The two control systems were each coupled separately with a micro-fission chamber to measure the current pulses of the neutron fluxes in the core at excess reactivity of 4 mk. The MCCLS and CC meter readings at a pre-set flux of 5.0 × 10{sup 11} n/cm{sup 2} s were 6.42 × 10{sup 11} n/cm{sup 2} s and 5.0 × 10{sup 11} n/cm{sup 2} s respectively. Due to ageing and obsolescence, the MCCLS and some components that control the sensitivity and the reading mechanism of the meters were replaced. One of the fission chambers was also removed and the two control systems were coupled to one fission chamber. The reliability of the neutronic parameters of the control systems was assessed after the replacement. The results showed that when the reactor is operated at different power levels of 5–30 kW using one micro-fission chamber, the pre-set neutron fluxes at the control systems is 1.6 times the neutron fluxes obtained using a flux monitor at the inner irradiation site two of the reactor. The average percentage deviations of the obtained fluxes from the pre-set values of 1.67 × 10{sup 11}–1.0 × 10{sup 12} n/cm{sup 2} s were 36.5%. This compares very well with the decrease in core excess reactivity of 36.3% of the nominal value of 4 mk, after operating the reactor at critical neutron flux of 1.0 × 10{sup 9} n/cm{sup 2} s.

  6. Calibration of the Failed-Fuel-Element Detection Systems in the Aagesta Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O.


    Results from a calibration of the systems for detection of fuel element ruptures in the Aagesta reactor are presented. The calibration was carried out by means of foils of zirconium-uranium alloy which were placed in a special fuel assembly. The release of fission products from these foils is due mainly to recoil and can be accurately calculated. Before the foils were used in the reactor their corrosion behaviour in high temperature water was investigated. The results obtained with the precipitator systems for bulk detection and localization are in good agreement with the expected performance. The sensitivity of these systems was found to be high enough for detection and localization of small defects of pin-hole type ({nu} = 10{sup -8}/s ). The general performance of the systems was satisfactory during the calibration tests, although a few adjustments are desirable. A bulk detecting system for monitoring of activities in the moderator, in which the {gamma}-radiation from coolant samples is measured directly after an ion exchanger, showed lower sensitivity than expected from calculations. It seems that the sensitivity of the latter system has to be improved to admit the detection of small defects. In the ion exchanger system, and to some extent in the precipitator systems, the background from A{sup 41} in the coolant limits the sensitivity. The calibration technique utilized seems to be of great advantage when investigating the performance of failed-fuel-element detection systems.

  7. A Project Management and Systems Engineering Structure for a Generation IV Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ed Gorski; Dennis Harrell; Finis Southworth


    The Very High Temperature Reactor (VHTR) will be an advanced, very high temperature (approximately 1000o C. coolant outlet temperature), gas cooled nuclear reactor and is the nearest term of six Generation IV reactor technologies for nuclear assisted hydrogen production. In 2001, the Generation IV International Forum (GIF), a ten nation international forum working together with the Department of Energy’s (DOE) Nuclear Energy Research Advisory Committee (NERAC), agreed to proceed with the development of a technology roadmap and identified the next generation of nuclear reactor systems for producing new sources of power. Since a new reactor has not been licensed in the United States since the 1970s, the risks are too large for a single utility to assume in the development of an unprecedented Generation IV reactor. The government must sponsor and invest in the research to resolve major first of a kind (FOAK) issues through a full-scale demonstration prior to industry implementation. DOE’s primary mission for the VHTR is to demonstrate nuclear reactor assisted cogeneration of electricity and hydrogen while meeting the Generation IV goals for safety, sustainability, proliferation resistance and physical security and economics. The successful deployment of the VHTR as a demonstration project will aid in restarting the now atrophied U.S. nuclear power industry infrastructure. It is envisioned that VHTR project participants will include DOE Laboratories, industry partners such as designers, constructors, manufacturers, utilities, and Generation IV international countries. To effectively mange R&D, engineering, procurement, construction, and operation for this multi-organizational and technologically complex project, systems engineering will be used extensively to ensure delivery of the final product. Although the VHTR is an unprecedented FOAK system, the R&D, when assessed using the Office of Science and Technology Gate Model, falls primarily in the 3rd - Exploratory

  8. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.


    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  9. Slaughterhouse wastewater treatment: evaluation of a new three-phase separation system in a UASB reactor. (United States)

    Caixeta, Cláudia E T; Cammarota, Magali C; Xavier, Alcina M F


    The anaerobic treatment of the wastewater from the meat processing industry was studied using a 7.2 1 UASB reactor. The reactor was equipped with an unconventional configuration of the three-phase separation system. The effluent was characterized in terms of pH (6.3-6.6), chemical oxygen demand (COD) (2,000-6,000 mg l(-1)), biochemical oxygen demand BOD5 (1,300-2,300 mg 1(-1)), fats (40-600 mg l(-1)) and total suspended solids (TSS) (850-6,300 mg l(-1)) The reactor operated continuously throughout 80 days with hydraulic retention time of 14, 18 and 22 h. The wastewater from Rezende Industrial was collected after it had gone through pretreatment (screening, flotation and equalization). COD, BOD and TSS reductions and the biogas production rate were the parameters considered in analyzing the efficiency of the process. The average production of biogas was 111 day(-1) (STP) for the three experimental runs. COD removal varied from 77% to 91% while BOD removal was 95%. The removal of total suspended solids varied from 81% to 86%. This fact supports optimal efficiency of the proposed three-phase separation system as well as the possibility of applying it to the treatment of industrial effluents.

  10. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)


    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  11. Study of DNA damage with a new system for irradiation of samples in a nuclear reactor. (United States)

    Gual, Maritza R; Milian, Felix M; Deppman, Airton; Coelho, Paulo R P


    In this paper, we report results of a quantitative analysis of the effects of neutrons on DNA, and, specifically, the production of simple and double breaks of plasmid DNA in aqueous solutions with different concentrations of free-radical scavengers. The radiation damage to DNA was evaluated by electrophoresis through agarose gels. The neutron and gamma doses were measured separately with thermoluminescent detectors. In this work, we have also demonstrated usefulness of a new system for positioning and removing samples in channel BH#3 of the IEA-R1 reactor at the Instituto de Pesquisas Energéticas e Nucleares (Brazil) without necessity of interrupting the reactor operation. Copyright © 2010 Elsevier Ltd. All rights reserved.

  12. Nuclear reactor fuel assembly duct-tube-to-inlet-nozzle attachment system (United States)

    Christiansen, David W.; Smith, Bob G.


    A reusable system for removably attaching the lower end 21 of a nuclear reactor fuel assembly duct tube to an upper end 11 of a nuclear reactor fuel assembly inlet nozzle. The duct tube's lower end 21 has sides terminating in locking tabs 22 which end in inwardly-extending flanges 23. The flanges 23 engage recesses 13 in the top section 12 of the inlet nozzle's upper end 11. A retaining collar 30 slides over the inlet nozzle's upper end 11 to restrain the flanges 23 in the recesses 13. A locking nut 40 has an inside threaded portion 41 which engages an outside threaded portion 15 of the inlet nozzle's upper end 11 to secure the retaining collar 30 against protrusions 24 on the duct tube's sides.

  13. A modular diagnosis system based on fuzzy logic for UASB reactors treating sewage. (United States)

    Borges, R M; Mattedi, A; Munaro, C J; Franci Gonçalves, R

    A modular diagnosis system (MDS), based on the framework of fuzzy logic, is proposed for upflow anaerobic sludge blanket (UASB) reactors treating sewage. In module 1, turbidity and rainfall information are used to estimate the influent organic content. In module 2, a dynamic fuzzy model is used to estimate the current biogas production from on-line measured variables, such as daily average temperature and the previous biogas flow rate, as well as the organic load. Finally, in module 3, all the information above and the residual value between the measured and estimated biogas production are used to provide diagnostic information about the operation status of the plant. The MDS was validated through its application to two pilot UASB reactors and the results showed that the tool can provide useful diagnoses to avoid plant failures.

  14. Passive containment cooling system with drywell pressure regulation for boiling water reactor (United States)

    Hill, Paul R.


    A boiling water reactor having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit.

  15. CFD Modeling of Flow and Ion Exchange Kinetics in a Rotating Bed Reactor System

    DEFF Research Database (Denmark)

    Larsson, Hilde Kristina; Schjøtt Andersen, Patrick Alexander; Byström, Emil


    A rotating bed reactor (RBR) has been modeled using computational fluid dynamics (CFD). The flow pattern in the RBR was investigated and the flow through the porous material in it was quantified. A simplified geometry representing the more complex RBR geometry was introduced and the simplified...... model was able to reproduce the main characteristics of the flow. Alternating reactor shapes were investigated, and it was concluded that the use of baffles has a very large impact on the flows through the porous material. The simulations suggested, therefore, that even faster reaction rates could...... be achieved by making the baffles deeper. Two-phase simulations were performed, which managed to reproduce the deflection of the gas–liquid interface in an unbaffled system. A chemical reaction was implemented in the model, describing the ion-exchange phenomena in the porous material using four different...

  16. Radiation safety assessment of a system of small reactors for distributed energy. (United States)

    Odano, N; Ishida, T


    A passively safe small reactor for a distributed energy system, PSRD, is an integral type of light-water reactor with a thermal output of 100 or 300 MW aimed to be used for supplying district heat, electricity to small grids, and so on. Candidate locations for the PSRD as a distributed energy source are on-ground, deep underground, and in a seaside pit in the vicinity of the energy consumption area. Assessments of the radiation safety of a PSRD were carried out for three cases corresponding to normal operation, shutdown and a hypothetical postulated accident for several siting candidates. Results of the radiation safety assessment indicate that the PSRD design has sufficient shielding performance and capability and that the exposure to the general public is very low in the case of a hypothetical accident.

  17. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao


    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  18. Reactor moderator, pressure vessel, and heat rejection system of an open-cycle gas core nuclear rocket concept (United States)

    Taylor, M. F.; Whitmarsh, C. L., Jr.; Sirocky, P. J., Jr.; Iwanczyke, L. C.


    A preliminary design study of a conceptual 6000-megawatt open-cycle gas-core nuclear rocket engine system was made. The engine has a thrust of 196,600 newtons (44,200 lb) and a specific impulse of 4400 seconds. The nuclear fuel is uranium-235 and the propellant is hydrogen. Critical fuel mass was calculated for several reactor configurations. Major components of the reactor (reflector, pressure vessel, and waste heat rejection system) were considered conceptually and were sized.

  19. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.


    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  20. Study of startup conditions of a pulsed annular reactor; Estudo das reacoes de partida de um reator anelar pulsado

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mario Augusto Bezerra da


    A new concept of reactor, which combines features of pulsed and stationary reactors, was proposed so as to produce intense neutronic fluxes. Such a reactor, known as VICHFPR (Very Intense Continuous High Flux Pulsed Reactor), consists of a subcritical core with an annular geometry and pulsed by a rotating reflector which acts as a reactivity modulator as it produces a short pulse (approximately equal to 1 ms) of high intensity, guiding the region near the pulser to super-prompt critical state. This dissertation intends to analyze the startup conditions of a Pulsed Annular Reactor. The evolution of the neutron pulse intensity is analyzed when the reactivity modulator is brought upwards according to a helicoidal path from its initial position (far away from the core), when the multiplication factor has a subcritical value, up to the final position (near the core), in which a super-prompt critical state is reached. Part of the analysis is based on the variation of neutron reflection, which is a uniform function of the exit and reflection angles between the core and the modulator. It must be emphasized that this work is an approximation of the real situation. As the initial and final reactor parameters are known, a programming code in Fortran is worked out to provide the multiplication factor and the flux intensity evolution. According to the results obtained with this code, the conditions under which the modulator must be lifted up during the startup are established. Basically, these conditions are related to the analysis of the rising and the rotation velocities, the reflector saving and the initial distance between the reactor and the modulator. The Pulsed Annular Reactor startup was divided into three stages. Because of its negative reactivity in the first two stages, the neutron multiplication is not large, while the last one, having a positive reactivity, shows an intense multiplication as is usually expected when handling pulsed systems. This last stage is quite

  1. Attenuation capability of low activation-modified high manganese austenitic stainless steel for fusion reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Eissa, M.M. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El-kameesy, S.U.; El-Fiki, S.A. [Physics Department, Faculty of Science, Ain Shams University, Cairo (Egypt); Ghali, S.N. [Steel Technology Department, Central Metallurgical Research and Development Institute (CMRDI), Helwan (Egypt); El Shazly, R.M. [Physics Department, Faculty of Science, Al-Azhar University, Cairo (Egypt); Saeed, Aly, E-mail: [Nuclear Power station Department, Faculty of Engineering, Egyptian-Russian University, Cairo (Egypt)


    Highlights: • Improvement stainless steel alloys to be used in fusion reactors. • Structural, mechanical, attenuation properties of investigated alloys were studied. • Good agreement between experimental and calculated results has been achieved. • The developed alloys could be considered as candidate materials for fusion reactors. - Abstract: Low nickel-high manganese austenitic stainless steel alloys, SSMn9Ni and SSMn10Ni, were developed to use as a shielding material in fusion reactor system. A standard austenitic stainless steel SS316L was prepared and studied as a reference sample. The microstructure properties of the present stainless steel alloys were investigated using Schaeffler diagram, optical microscopy, and X-ray diffraction pattern. Mainly, an austenite phase was observed for the prepared stainless steel alloys. Additionally, a small ferrite phase was observed in SS316L and SSMn10Ni samples. The mechanical properties of the prepared alloys were studied using Vickers hardness and tensile tests at room temperature. The studied manganese stainless steel alloys showed higher hardness, yield strength, and ultimate tensile strength than SS316L. On the other hand, the manganese stainless steel elongation had relatively lower values than the standard SS316L. The removal cross section for both slow and total slow (primary and those slowed down in sample) neutrons were carried out using {sup 241}Am-Be neutron source. Gamma ray attenuation parameters were carried out for different gamma ray energy lines which emitted from {sup 60}Co and {sup 232}Th radioactive sources. The developed manganese stainless steel alloys had a higher total slow removal cross section than SS316L. While the slow neutron and gamma rays were nearly the same for all studied stainless steel alloys. From the obtained results, the developed manganese stainless steel alloys could be considered as candidate materials for fusion reactor system with low activation based on the short life

  2. Development of automatic reactor vessel inspection systems: development of data acquisition and analysis system for the nuclear vessel weld

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. H.; Lim, H. T.; Um, B. G. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)


    The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine the reactor vessel weldsIn order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed in this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition and analysis software was developed. 11 refs., 6 figs., 9 tabs. (Author)

  3. Instrumentation and control system for the prototype fast breeder reactor 'MONJU' power station

    Energy Technology Data Exchange (ETDEWEB)

    Hara, Hiroshi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)); Mae, Yoshinori; Ishida, Takayuki; Hashiura, Kazuhiko; Kasai, Shozo; Yamamoto, Hajime


    The fast breeder reactor 'Monju' power station is constructed as the nuclear power station of next generation in Tsuruga City, Fukui Prefecture. In order to realize high safety and operational reliability as the newest nuclear power station, the measurement and control system of Monju (electric power output 280 MW) has been designed and manufactured by reflecting the experiences of construction and operation of the experimental FBR 'Joyo' and the results of various research and development of sodium instrumentation and others, and by using the latest digital control technology and multiplexing system technology. In this paper, the results of development of the characteristic measurement and control technology as fast breeder reactors and the state of application to the measurement and control system which was designed and manufactured for Monju are described. Central monitoring panel, plant control system, sodium instrumentation, preheating control system and so on are reported. In the case of Monju, the heat capacity and thermal inertia of the primary and secondary cooling systems are large, and the system comprises three loops. (K.I.).

  4. Application of S-CO{sub 2} Cycle for Small Modular Reactor coupled with Desalination System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Woong; Bae, Seong Jun; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)


    The Korean small modular reactor, SMART (System-integrated Modular Advanced ReacTor, 100MWe), is designed to achieve enhanced safety and improved economics through reliable passive safety systems, a system simplification and component modularization. SMART can generate electricity and provide water by seawater desalination. However, due to the desalination aspect of SMART, the total amount of net electricity generation is decreased from 100MWe to 90MWe. The authors suggest in this presentation that the reduction of electricity generation can be replenished by applying S-CO{sub 2} power cycle technology. The S-CO{sub 2} Brayton cycle, which is recently receiving significant attention as the next generation power conversion system, has some benefits such as high cycle efficiency, simple configuration, compactness and so on. In this study, the cycle performance analysis of the S-CO{sub 2} cycles for SMART with desalination system is conducted. The simple recuperated S-CO{sub 2} cycle is revised for coupling with desalination system. The three revised layout are proposed for the cycle performance comparison. In this results of the 3rd revised layout, the cycle efficiency reached 37.8%, which is higher than the efficiency of current SMART with the conventional power conversion system 30%.

  5. Experimental Study of Subcritical Water Liquefaction of Biomass

    DEFF Research Database (Denmark)

    Zhu, Zhe; Toor, Saqib; Rosendahl, Lasse


    In this work, hydrothermal liquefaction (HTL) of wood industry residues (wood, bark, sawdust) and macroalgae for producing biofuels has been investigated under subcritical water conditions (at temperature of 300 C), with and without the presence of catalyst. The effects of catalyst and biomass type...

  6. Local energy losses at positive and negative steps in subcritical ...

    African Journals Online (AJOL)

    Local energy losses occur when there is a transition in open channel flow. Even though local losses in subcritical open channel flow due to changes in channel width have been studied, to date no studies have been reported for losses due to changes in bed elevations. Steps are commonly used in engineering applications ...

  7. Local energy losses at positive and negative steps in subcritical ...

    African Journals Online (AJOL)


    7) 554-568. MORRIS HM and WIGGERT JM (1972) Applied Hydraulics in. Engineering. John Wiley & Sons, New York. ÖRSEL SI (2002) Local Losses at a Step in a Sub-critical Open. Channel Flow. M.Sc. Thesis, Department ...

  8. Extraction of antioxidants from Chlorella sp. using subcritical water treatment (United States)

    Zakaria, S. M.; Mustapa Kamal, S. M.; Harun, M. R.; Omar, R.; Siajam, S. I.


    Chlorella sp. microalgae is one of the main source of natural bioactive compounds used in the food and pharmaceutical industries. Subcritical water extraction is the technique that offers an efficient, non-toxic, and environmental-friendly method to obtain natural ingredients. In this work, the extracts of Chlorella sp. microalgae was evaluated in terms of: chemical composition, extraction (polysaccharides) yield and antioxidant activity, using subcritical water extraction. Extractions were performed at temperatures ranging from 100°C to 300°C. The results show that by using subcritical water, the highest yield of polysaccharides is 23.6 that obtained at 150°C. Analysis on the polysaccharides yield show that the contents were highly influenced by the extraction temperature. The individual antioxidant activity were evaluated by in vitro assay using a free radical method. In general, the antioxidant activity of the extracts obtained at different water temperatures was high, with values of 31.08-54.29 . The results indicated that extraction by subcritical water was effective and Chlorella sp. can be a useful source of natural antioxidants.

  9. Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

    Energy Technology Data Exchange (ETDEWEB)

    Kunsman, David Marvin; Aldemir, Tunc (Ohio State University); Rutt, Benjamin (Ohio State University); Metzroth, Kyle (Ohio State University); Catalyurek, Umit (Ohio State University); Denning, Richard (Ohio State University); Hakobyan, Aram (Ohio State University); Dunagan, Sean C.


    This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accident progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other

  10. Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I


    One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

  11. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady


    Full Text Available The paper presents the calculated dose rate carried out by using MicroShield code to show the importance of hot water layer system (HWL in 22 MW open pool type reactor from the radiation protection safety point of view. The paper presents the dose rate profiles over the pool surface in normal and abnormal operations of HWL system. The results show that, in case of losing the hot water layer effect, the radiation dose rate profiles over the pool surface will increase from values lower than the worker permissible dose limits to values very higher than the permissible dose limits.

  12. Review of the Tri-Agency Space Nuclear Reactor Power System Technology Program (United States)

    Ambrus, J. H.; Wright, W. E.; Bunch, D. F.


    The Space Nuclear Reactor Power System Technology Program designated SP-100 was created in 1983 by NASA, the U.S. Department of Defense, and the Defense Advanced Research Projects Agency. Attention is presently given to the development history of SP-100 over the course of its first year, in which it has been engaged in program objectives' definition, the analysis of civil and military missions, nuclear power system functional requirements' definition, concept definition studies, the selection of primary concepts for technology feasibility validation, and the acquisition of initial experimental and analytical results.

  13. Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin


    Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

  14. Nitrifying-denitrifying filters and UV-C disinfection reactor: a combined system for wastewater treatment. (United States)

    Ben Rajeb, Asma; Mehri, Inès; Nasr, Houda; Najjari, Afef; Saidi, Neila; Hassen, Abdennaceur


    Biological treatment systems use the natural processes of ubiquitous organisms to remove pollutants and improve the water quality before discharge to the environment. In this paper, the nitrification/denitrification reactor allowed a reduction in organic load, but offered a weak efficiency in nitrate reduction. However, the additions of the activated sludge in the reactor improve this efficiency. A decrease of [Formula: see text] values from 13.3 to 8 mg/l was noted. Nevertheless, sludge inoculation led to a net increase of the number of pathogenic bacteria. For this reason, a UV-C pilot reactor was installed at the exit of the biological nitrification-denitrification device. Thus, a fluence of 50 was sufficient to achieve values of 20 MPN/100 ml for fecal coliform and 6 MPN/100 ml for fecal streptococci, conforms to Tunisian Standards of Rejection. On the other hand, the DGGE approach has allowed a direct assessment of the bacterial community changes upon the treated wastewater.

  15. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Borum, Robert C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chaleff, Ethan S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rogerson, Doug W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J. [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation, Canton, MI (United States)


    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  16. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Avincola, V.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission Joint Research Centre - Institute for Transuranium Elements (JRC-ITU) (Germany)


    The High Temperature Reactor (HTR) is an advanced reactor concept with particular safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (tri-isotropic) coating designed to provide high fission product retention. Passive safety features of the HTR include a low power density in the core compared to other reactor designs; this ensures sufficient heat transport in a loss of coolant accident scenario. The temperature during such events would not exceed 1600 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under severe accident conditions is necessary to confirm the fission product retention of TRISO coated particles and to validate relevant computer codes. Though helium is used as coolant for the HTR system, additional corrosion effects come into play in case of an in-leakage affecting the primary circuit. The experimental scope of the present work focuses on two key aspects associated with the HTR fuel safety. Fission product retention at high temperatures (up to {proportional_to}1800 C) is analyzed with the so-called cold finger apparatus (KueFA: Kuehlfinger-Apparatur), while the performance of HTR fuel elements in case of air/steam ingress accidents is assessed with a high temperature corrosion apparatus (KORA: Korrosions-Apparatur). (orig.)

  17. Integrated photocatalytic and sequencing batch reactor (SBR) treatment system for degradation of phenol (United States)

    Yusoff, Nik Noor Athirah Nik; Ong, Soon-An; Ho, Li-Ngee; Wong, Yee-Shian; Khalik, Wan Fadhilah Wan Mohd; Lee, Sin-Li


    This study will examine the efficiency of the simultaneous photocatalytic and biodegradation process in the same treatment reactor. The sequencing batch reactor or also known as SBR is an effective wastewater treatment method that has been applied widely. SBR system has become an alternative method for industrial wastewater treatment with high concentration of chemical oxygen demand (COD), and phenolic compound. In order for the photocatalytic process to occur, ZnO nanoparticles immobilized onto sponge were introduced to the reactor. It was observed that the COD value were decreased, indicated that the simultaneous biodegradation and photodegradation process in functional. The effect of ZnO nanoparticles on the production and composition of extracellular polymeric substances (EPS) and the physiochemical stability of activated sludge in hybrid growth type SBR were monitored. The percentages of removal are varied with different concentration of ZnO nanoparticles. The highest COD removal recorded is 31.5% with concentration of ZnO 0.6 mg/L. With the present of the ZnO nanoparticles, the degradation of phenol was relatively better than combination of biological of photlysis and biological.

  18. Development of Integrated Regulatory Aging Management System related to Reactor Vessel Internals

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Hanok; Park, Jeongsoon; Kim, Seonjae; Jhung, Myungjo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)


    The primary function of the reactor vessel internals (RVIs) is to support the core, the control rod assemblies, the core support structure and the reactor pressure vessel (RPV) surveillance capsules. The RVIs have the additional function to direct the flow of the reactor coolant and provide shielding for the RPV. Ageing mechanisms are specific processes that gradually change characteristics of a component with time and use. According to the Generic Aging Lessons Learned (GALL) report, aging mechanisms, such as fatigue, embrittlement, corrosion, wear, radiation induced creep, relaxation and swelling, is related to RVIs. Establishing that effects of aging degradation in RVIs are adequately managed is vital for assuring continued functionality of RVIs. To achieve this goal, it is necessary to develop the regulatory standard as well as generic inspection and evaluation guideline for RVIs. In this paper, the Integrated Regulatory Aging Management System (IR-Aging), which efficiently manages key data necessary to the development of regulatory standards and assists effective evaluation of RVIs, is proposed. By using the proposed system, experts in different fields can co-operate to resolve safety issues and all users can share information and create valuable knowledge-base. In this paper, the Integrated Regulatory Aging Management System (IR-Aging) is proposed in order to manage data necessary to the development of regulatory standards and assists effective evaluation of RVIs. The proposed system provides various documents, such as US NRC and domestic regulatory documents, licensee's documents submitted to a regulatory body, and research documents. By using the proposed system, experts in different fields can co-operate to resolve safety issues and all users can share information and create valuable knowledge-base.

  19. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Professor Neill Todreas


    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

  20. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux. (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A


    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

  1. (I) A Declarative Framework for ERP Systems(II) Reactors: A Data-Driven Programming Model for Distributed Applications

    DEFF Research Database (Denmark)

    Stefansen, Christian Oskar Erik

    . • Using Soft Constraints to Guide Users in Flexible Business Process Management Systems. The paper shows how the inability of a process language to express soft constraints—constraints that can be violated occasionally, but are closely monitored—leads to a loss of intentional information in process....../Asynchronous Programming Model for Distributed Applications. The paper motivates, explains, and defines a distributed data-driven programming model. In the model a reactor is a stateful unit of distribution. A reactor specifies constructive, declarative constraints on its data and the data of other reactors in the style...

  2. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy


    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  3. Training courses on neutron detection systems on the ISIS research reactor: on-site and through internet training

    Energy Technology Data Exchange (ETDEWEB)

    Lescop, B.; Badeau, G.; Ivanovic, S.; Foulon, F. [National Institute for Nuclear science and Technology French Atomic Energy and Alternative Energies Commission (CEA), Saclay Research Center, 91191 Gif-sur-Yvette (France)


    Today, ISIS research reactor is an essential tool for Education and Training programs organized by the National Institute for Nuclear Science and Technology (INSTN) from CEA. In the field of nuclear instrumentation, the INSTN offers both, theoretical courses and training courses on the use of neutron detection systems taking advantage of the ISIS research reactor for the supply of a wide range of neutron fluxes. This paper describes the content of the training carried out on the use of neutron detectors and detection systems, on-site or remote. The ISIS reactor is a 700 kW open core pool type reactor. The facility is very flexible since neutron detectors can be inserted into the core or its vicinity, and be used at different levels of power according to the needs of the course. Neutron fluxes, typically ranging from 1 to 10{sup 12} n/cm{sup 2}.s, can be obtained for the characterisation of the neutron detectors and detection systems. For the monitoring of the neutron density at low level of power, the Instrumentation and Control (I and C) system of the reactor is equipped with two detection systems, named BN1 and BN2. Each way contains a fission chamber, type CFUL01, connected to an electronic system type SIREX.The system works in pulse mode and exhibits two outputs: the counting rate and the doubling time. For the high level of power, the I and C is equipped with two detection systems HN1 and HN2.Each way contain a boron ionization chamber (type CC52) connected to an electronics system type SIREX. The system works in current mode and has two outputs: the current and the doubling time. For each mode, the trainees can observe and measure the signal at the different stages of the electronic system, with an oscilloscope. They can understand the role of each component of the detection system: detector, cable and each electronic block. The limitation of the detection modes and their operating range can be established from the measured signal. The trainees can also

  4. Semiclassical Limit of the Non-linear Schroedinger-Poisson Equation With Subcritical Initial Data (United States)


    lim ∇xargψ. As noted earlier, this argument is self - consistent as long as the solution of the Euler- Poisson system (1.5)-(1.6) remains classical...00-2003 to 00-00-2003 4. TITLE AND SUBTITLE Semiclassical Limit of the Non-linear Schrodinger - Poisson Equation with Subcritical Initial Data 5a...classical limit of a self - consistent quantum-Vlasov equation in 3-D, Math. Models Methods Appl. Sci., 3 (1993), pp. 109–124. [SMM] C. Sparber, P. Markowich

  5. Chemical looping reactor system design double loop circulating fluidized bed (DLCFB)

    Energy Technology Data Exchange (ETDEWEB)

    Bischi, Aldo


    Chemical looping combustion (CLC) is continuously gaining more importance among the carbon capture and storage (CCS) technologies. It is an unmixed combustion process which takes place in two steps. An effective way to realize CLC is to use two interconnected fluidized beds and a metallic powder circulating among them, acting as oxygen carrier. The metallic powder oxidizes at high temperature in one of the two reactors, the air reactor (AR). It reacts in a highly exothermic reaction with the oxygen of the injected fluidising air. Afterwards the particles are sent to the other reactor where the fuel is injected, the fuel reactor (FR). There, they transport heat and oxygen necessary for the reaction with the injected fuel to take place. At high temperatures, the particle's oxygen reacts with the fuel producing Co2 and steam, and the particles are ready to start the loop again. The overall reaction, the sum of the enthalpy changes of the oxygen carrier oxidation and reduction reactions, is the same as for the conventional combustion. Two are the key features, which make CLC promising both for costs and capture efficiency. First, the high inherent irreversibility of the conventional combustion is avoided because the energy is utilized stepwise. Second, the Co2 is intrinsically separated within the process; so there is in principle no need either of extra carbon capture devices or of expensive air separation units to produce oxygen for oxy-combustion. A lot of effort is taking place worldwide on the development of new chemical looping oxygen carrier particles, reactor systems and processes. The current work is focused on the reactor system: a new design is presented, for the construction of an atmospheric 150kWth prototype working with gaseous fuel and possibly with inexpensive oxygen carriers derived from industrial by-products or natural minerals. It consists of two circulating fluidized beds capable to operate in fast fluidization regime; this will increase the

  6. Detailed Design of the Safety Residual Heat Removal System and a Circulation Pump for the KIJANG Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyungi; Seo, Kyoungwoo; Kim, Seonghoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Primary cooling system (PCS) circulates the coolant from the reactor core to the heat exchanger. Therefore the heat generated from the fuel assembly in the reactor core is removed continuously. The PCS is designed based on the required thermal design flow rate of the reactor core, uncertainty of measuring instruments and the safe functions. Primary coolant is generally dumped into the pool and goes to the reactor core through the flow guide. The fission heat generated from the fuel assembly is transferred to the coolant, and then heated coolant goes to the PCS equipment room in order to remove the heat through the heat exchanger. SRHSR is designed based on the required flow rate and system constraints. Centrifugal pump of Case 1 with a non-dimensional specific speed of 0.97 [-] and specific diameter of 3.33 [-] is chosen as the SRHRS pump for the KJRR.


    Directory of Open Access Journals (Sweden)

    Anna Hajduk


    Full Text Available An alternative to aerobic wastewater treatment systems are anaerobic reactors. When designing anaerobic reactors attention is paid to the appropriate filling, pumping systems, or mixing systems, enabling the re-duction of technological limitations, which contribute to the improvement of end effects such as, quantity and quality of the resulting biogas and the quality of treated wastewater. Described experiment related to researches on the evaluation of the efficiency of removing contamina-tions from synthetic dairy waste water using anaerobic reactor equipped with an innovative mixing sys-tem. The efficiency of removal of organic compounds made studies ranged from 96.25% to 99.03%. The concentration of total nitrogen in raw wastewater was at a level of 148.36 ± 0 mg N/dm3 to 593.42 ± 94,92 mg N/dm3, and treated wastewater from 21.66 ± 19.71 mg N/dm3 to 28.73 ± 0.4 mg N/dm3. The concentration of total phosphorus in raw wastewater was at a level of 110 ± 0 mg P/dm3 to 441.16 ± 19.83 mg P/dm3, and treated wastewater from 16.49 ± 16.13 mg P/dm3 to 354 ± 14.18 mg P/dm3. The methane content of the biogas produced was at a level of from 0.0413 dm3 per 1 g COD introduced to 0.4367 dm3 per 1 g COD introduced.

  8. Concept of turbines for ultrasupercritical, supercritical, and subcritical steam conditions (United States)

    Mikhailov, V. E.; Khomenok, L. A.; Pichugin, I. I.; Kovalev, I. A.; Bozhko, V. V.; Vladimirskii, O. A.; Zaitsev, I. V.; Kachuriner, Yu. Ya.; Nosovitskii, I. A.; Orlik, V. G.


    The article describes the design features of condensing turbines for ultrasupercritical initial steam conditions (USSC) and large-capacity cogeneration turbines for super- and subcritical steam conditions having increased steam extractions for district heating purposes. For improving the efficiency and reliability indicators of USSC turbines, it is proposed to use forced cooling of the head high-temperature thermally stressed parts of the high- and intermediate-pressure rotors, reaction-type blades of the high-pressure cylinder (HPC) and at least the first stages of the intermediate-pressure cylinder (IPC), the double-wall HPC casing with narrow flanges of its horizontal joints, a rigid HPC rotor, an extended system of regenerative steam extractions without using extractions from the HPC flow path, and the low-pressure cylinder's inner casing moving in accordance with the IPC thermal expansions. For cogeneration turbines, it is proposed to shift the upper district heating extraction (or its significant part) to the feedwater pump turbine, which will make it possible to improve the turbine plant efficiency and arrange both district heating extractions in the IPC. In addition, in the case of using a disengaging coupling or precision conical bolts in the coupling, this solution will make it possible to disconnect the LPC in shifting the turbine to operate in the cogeneration mode. The article points out the need to intensify turbine development efforts with the use of modern methods for improving their efficiency and reliability involving, in particular, the use of relatively short 3D blades, last stages fitted with longer rotor blades, evaporation techniques for removing moisture in the last-stage diaphragm, and LPC rotor blades with radial grooves on their leading edges.

  9. Radioactive isotope production for medical applications using Kharkov electron driven subcritical assembly facility.

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, A.; Gohar, Y.; Nuclear Engineering Division


    Kharkov Institute of Physics and Technology (KIPT) of Ukraine has a plan to construct an accelerator driven subcritical assembly. The main functions of the subcritical assembly are the medical isotope production, neutron thereby, and the support of the Ukraine nuclear industry. Reactor physics experiments and material research will be carried out using the capabilities of this facility. The United States of America and Ukraine have started collaboration activity for developing a conceptual design for this facility with low enrichment uranium (LEU) fuel. Different conceptual designs are being developed based on the facility mission and the engineering requirements including nuclear physics, neutronics, heat transfer, thermal hydraulics, structure, and material issues. Different fuel designs with LEU and reflector materials are considered in the design process. Safety, reliability, and environmental considerations are included in the facility conceptual design. The facility is configured to accommodate future design improvements and upgrades. This report is a part of the Argonne National Laboratory Activity within this collaboration for developing and characterizing the subcritical assembly conceptual design. In this study, the medical isotope production function of the Kharkov facility is defined. First, a review was carried out to identify the medical isotopes and its medical use. Then a preliminary assessment was performed without including the self-shielding effect of the irradiated samples. Finally, more detailed investigation was carried out including the self-shielding effect, which defined the sample size and irradiation location for producing each medical isotope. In the first part, the reaction rates were calculated as the multiplication of the cross section with the unperturbed neutron flux of the facility. Over fifty isotopes were considered and all transmutation channels are used including (n,{gamma}), (n,2n), (n,p), and ({gamma},n). In the second part

  10. Efficient, high-speed methane fermentation for sewage sludge using subcritical water hydrolysis as pretreatment. (United States)

    Yoshida, Hiroyuki; Tokumoto, Hayato; Ishii, Kyoko; Ishii, Ryo


    A novel biomass-energy process for the production of methane from sewage sludge using a subcritical water (sub-CW) hydrolysis reaction as pretreatment is proposed. The main substances of sewage sludge hydrolyzed by sub-CW at 513 K for 10 min were acetic acid, formic acid, pyroglutamic acid, alanine, and glycine. Fermentation experiments were conducted in an anaerobic-sludge reactor for two different samples: real sewage sludge and a model solution containing components typically produced by the sub-CW pretreatment of sewage sludge. In the experiment for the sub-CW pretreatment of sewage sludge, methane generation was twice that for non-pretreatment after 3 days of incubation. In the model experiment, the methane conversion was about 40% with the application of mixture of organic acids and amino acids after 5 days of incubation. Furthermore, the methane conversion was about 60% for 2 days when only organic acids, such as acetic acid and formic acid, were applied. Because acetic acid is the key intermediate and main precursor of the methanogenesis step, fermentation experiments were conducted in an anaerobic-sludge reactor with high concentrations of acetic acid (0.01-0.1M). Nearly 100% of acetic acid was converted to methane and carbon dioxide in 1-3 days.

  11. Elements of record management system for the RA research reactor decommissioning

    Directory of Open Access Journals (Sweden)

    Stejić Milijana


    Full Text Available According to latest recommendations, the record management system of a nuclear facility should operate as a part of the integrated management information system, and is implemented at the very beginning of the facility’s life cycle. The record management becomes particularly important at the end of the operation of a facility and then the operational record management system gradually transforms to a decommissioning one. However there is a significant number of nuclear facilities in the world which have reached the decommissioning stage with out having neither the initial decommissioning plan nor the established record management system. The objective of this paper is to introduce constituted elements of the record management system for the decommissioning of the RA research reactor in the VINČA Institute of Nuclear Sciences, and to discuss future planned actions related to this matter.

  12. Systems and methods for harvesting and storing materials produced in a nuclear reactor (United States)

    Heinold, Mark R.; Dayal, Yogeshwar; Brittingham, Martin W.


    Systems produce desired isotopes through irradiation in nuclear reactor instrumentation tubes and deposit the same in a robust facility for immediate shipping, handling, and/or consumption. Irradiation targets are inserted and removed through inaccessible areas without plant shutdown and placed in the harvesting facility, such as a plurality of sealable and shipping-safe casks and/or canisters. Systems may connect various structures in a sealed manner to avoid release of dangerous or unwanted matter throughout the nuclear plant, and/or systems may also automatically decontaminate materials to be released. Useable casks or canisters can include plural barriers for containment that are temporarily and selectively removable with specially-configured paths inserted therein. Penetrations in the facilities may limit waste or pneumatic gas escape and allow the same to be removed from the systems without over-pressurization or leakage. Methods include processing irradiation targets through such systems and securely delivering them in such harvesting facilities.

  13. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.


    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  14. Closed Brayton Cycle power system with a high temperature pellet bed reactor heat source for NEP applications (United States)

    Juhasz, Albert J.; El-Genk, Mohamed S.; Harper, William B., Jr.


    Capitalizing on past and future development of high temperature gas reactor (HTGR) technology, a low mass 15 MWe closed gas turbine cycle power system using a pellet bed reactor heating helium working fluid is proposed for Nuclear Electric Propulsion (NEP) applications. Although the design of this directly coupled system architecture, comprising the reactor/power system/space radiator subsystems, is presented in conceptual form, sufficient detail is included to permit an assessment of overall system performance and mass. Furthermore, an attempt is made to show how tailoring of the main subsystem design characteristics can be utilized to achieve synergistic system level advantages that can lead to improved reliability and enhanced system life while reducing the number of parasitic load driven peripheral subsystems.

  15. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo


    Full Text Available Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipments. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogenously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occured in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, ekspansi load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (rerouting is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node

  16. Generation IV: new reactor systems; Neue Reaktorsysteme innerhalb der Generation IV Initiative

    Energy Technology Data Exchange (ETDEWEB)

    Starflinger, J.; Schulenberg, T. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). IKET; Hofmeister, J. [RWE Power AG, Regenerative Stromerzeugung, Essen (Germany); Tromm, W. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Programm Nukleare Sicherheitsforschung


    Generation IV, an initiative for international cooperation in nuclear technology, was launched by 10 states in 2000 and joined by Euratom in July 2003. Its aim is to assess nuclear energy systems complying with future safety, disposal, proliferation, and public acceptance requirements. The Forschungszentrum Karlsruhe focuses on design, thermohydraulics, and neutron kinetics. Work is mainly devoted to the high-performance light water reactor (HPLWR) with supercritical steam conditions. Thus, competence can be maintained, as the HPLWR issues qualify for later work in nuclear industry. (orig.)


    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick


    Key requirements that affect the design of the high temperature gas-cooled reactor nuclear heat supply system (HTGR-NHSS) as the NGNP Project progresses through the design, licensing, construction and testing of the first of a kind HTGR based plant are summarized. These requirements derive from pre-conceptual design development completed to-date by HTGR Suppliers, collaboration with potential end users of the HTGR technology to identify energy needs, evaluation of integration of the HTGR technology with industrial processes and recommendations of the NGNP Project Senior Advisory Group.

  18. ENFORM II: a calculational system for light water reactor logistics and effluent analysis

    Energy Technology Data Exchange (ETDEWEB)

    Heeb, C.M.; Lewallen, M.A.; Purcell, W.L.; Cole, B.M.


    ENFORM is a computer-based information system that addresses the material logistics, environmental releases and economics of light water reactor (LWR) operation. The most important system inputs consist of electric energy generation requirements, details of plant construction scheduling, unit costs, and environmental release factors. From these inputs the ENFORM system computes the mass balances and generates the environmental release information for noxious chemicals and radionuclides from various fuel cycle facilities (except waste disposal). Fuel cycle costs and electric power costs are also computed. All code development subsequent to 1977 is summarized. Programming instructions are provided for the modules that are comprised in the ENFORM system. ENGEN, a code that uses a generation schedule specified by the user and isotopic data generated by ORIGEN, has been developed to produce a scenario-specific data base. Other codes (ENMAT, ENRAD, etc) have been developed to use data base information to estimate radioactive and nonradioactive release information.

  19. Analysis of closed cycle megawatt class space power systems with nuclear reactor heat sources (United States)

    Juhasz, A. J.; Jones, B. I.


    The analysis and integration studies of multimegawatt nuclear power conversion systems for potential SDI applications is presented. A study is summarized which considered 3 separate types of power conversion systems for steady state power generation with a duty requirement of 1 yr at full power. The systems considered are based on the following conversion cycles: direct and indirect Brayton gas turbine, direct and indirect liquid metal Rankine, and in core thermionic. A complete mass analysis was performed for each system at power levels ranging from 1 to 25 MWe for both heat pipe and liquid droplet radiator options. In the modeling of common subsystems, reactor and shield calculations were based on multiparameter correlation and an in-house analysis for the heat rejection and other subsystems.

  20. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar


    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  1. Successional development of biofilms in moving bed biofilm reactor (MBBR) systems treating municipal wastewater. (United States)

    Biswas, Kristi; Taylor, Michael W; Turner, Susan J


    Biofilm-based technologies, such as moving bed biofilm reactor (MBBR) systems, are widely used to treat wastewater. Biofilm development is important for MBBR systems as much of the microbial biomass is retained within reactors as biofilm on suspended carriers. Little is known about this process of biofilm development and the microorganisms upon which MBBRs rely. We documented successional changes in microbial communities as biofilms established in two full-scale MBBR systems treating municipal wastewater over two seasons. 16S rRNA gene-targeted pyrosequencing and clone libraries were used to describe microbial communities. These data indicate a successional process that commences with the establishment of an aerobic community dominated by Gammaproteobacteria (up to 52 % of sequences). Over time, this community shifts towards dominance by putatively anaerobic organisms including Deltaproteobacteria and Clostridiales. Significant differences were observed between the two wastewater treatment plants (WWTPs), mostly due to a large number of sequences (up to 55 %) representing Epsilonproteobacteria (mostly Arcobacter) at one site. Archaea in young biofilms included several lineages of Euryarchaeota and Crenarchaeota. In contrast, the mature biofilm consisted entirely of Methanosarcinaceae (Euryarchaeota). This study provides new insights into the community structure of developing biofilms at full-scale WWTPs and provides the basis for optimizing MBBR start-up and operational parameters.

  2. Design and development of fast pneumatic transfer system (PTS) for instrumental neutron activation analysis at Jordan research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yongsam; Kim, Sunha; Moon, Jonghwa; Choi, Jinbok; Lee, Jongmin; Ryu, Jungsu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    A pneumatic transfer system (PTS) is one of the important equipment used for an neutron irradiation of a target material for an instrumental neutron activation analysis (INAA) in a research reactor. In particular, a rapid pneumatic transportation of irradiation capsule is essential for an accurate measurement of a short half-life nuclide. Three types of PTS for NAA facility at the Jordan Research and Training Reactor (JRTR) were newly developed for a functional improvement involving a manual and an automatic system which is equipped with programmable logic controller, software, and 13 devices to facilitate optimal operation of the system. In this paper, the designs and construction of these PTS, the operation and control of the system are described. In addition, a functional and operational test of the system were carried out as one of the basic requirement and characteristic parameters, and the results were reported to provide a user information as well as for the management and safety of the reactor.

  3. Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau


    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

  4. The influence of pH adjustment on kinetics parameters in tapioca wastewater treatment using aerobic sequencing batch reactor system (United States)

    Mulyani, Happy; Budianto, Gregorius Prima Indra; Margono, Kaavessina, Mujtahid


    The present investigation deals with the aerobic sequencing batch reactor system of tapioca wastewater treatment with varying pH influent conditions. This project was carried out to evaluate the effect of pH on kinetics parameters of system. It was done by operating aerobic sequencing batch reactor system during 8 hours in many tapioca wastewater conditions (pH 4.91, pH 7, pH 8). The Chemical Oxygen Demand (COD) and Mixed Liquor Volatile Suspended Solids (MLVSS) of the aerobic sequencing batch reactor system effluent at steady state condition were determined at interval time of two hours to generate data for substrate inhibition kinetics parameters. Values of the kinetics constants were determined using Monod and Andrews models. There was no inhibition constant (Ki) detected in all process variation of aerobic sequencing batch reactor system for tapioca wastewater treatment in this study. Furthermore, pH 8 was selected as the preferred aerobic sequencing batch reactor system condition in those ranging pH investigated due to its achievement of values of kinetics parameters such µmax = 0.010457/hour and Ks = 255.0664 mg/L COD.

  5. Ion-scale turbulence in MAST: anomalous transport, subcritical transitions, and comparison to BES measurements (United States)

    van Wyk, F.; Highcock, E. G.; Field, A. R.; Roach, C. M.; Schekochihin, A. A.; Parra, F. I.; Dorland, W.


    We investigate the effect of varying the ion temperature gradient (ITG) and toroidal equilibrium scale sheared flow on ion-scale turbulence in the outer core of MAST by means of local gyrokinetic simulations. We show that nonlinear simulations reproduce the experimental ion heat flux and that the experimentally measured values of the ITG and the flow shear lie close to the turbulence threshold. We demonstrate that the system is subcritical in the presence of flow shear, i.e., the system is formally stable to small perturbations, but transitions to a turbulent state given a large enough initial perturbation. We propose that the transition to subcritical turbulence occurs via an intermediate state dominated by low number of coherent long-lived structures, close to threshold, which increase in number as the system is taken away from the threshold into the more strongly turbulent regime, until they fill the domain and a more conventional turbulence emerges. We show that the properties of turbulence are effectively functions of the distance to threshold, as quantified by the ion heat flux. We make quantitative comparisons of correlation lengths, times, and amplitudes between our simulations and experimental measurements using the MAST BES diagnostic. We find reasonable agreement of the correlation properties, most notably of the correlation time, for which significant discrepancies were found in previous numerical studies of MAST turbulence.

  6. Design of a Rail Gun System for Mitigating Disruptions in Fusion Reactors (United States)

    Lay, Wei-Siang

    Magnetic fusion devices, such as the tokamak, that carry a large amount of current to generate the plasma confining magnetic fields have the potential to lose magnetic stability control. This can lead to a major plasma disruption, which can cause most of the stored plasma energy to be lost to localized regions on the walls, causing severe damage. This is the most important issue for the $20B ITER device (International Thermonuclear Experimental Reactor) that is under construction in France. By injecting radiative materials deep into the plasma, the plasma energy could be dispersed more evenly on the vessel surface thus mitigating the harmful consequences of a disruption. Methods currently planned for ITER rely on the slow expansion of gases to propel the radiative payloads, and they also need to be located far away from the reactor vessel, which further slows down the response time of the system. Rail guns are being developed for aerospace applications, such as for mass transfer from the surface of the moon and asteroids to low earth orbit. A miniatured version of this aerospace technology seems to be particularly well suited to meet the fast time response needs of an ITER disruption mitigation system. Mounting this device close to the reactor vessel is also possible, which substantially increases its performance because the stray magnetic fields near the vessel walls could be used to augment the rail gun generated magnetic fields. In this thesis, the potential viability on Rail Gun based DMS is studied to investigate its projected fast time response capability by design, fabrication, and experiment of an NSTX-U sized rail gun system. Material and geometry based tests are used to find the most suitable armature design for this system for which the desirable attributes are high specific stiffness and high electrical conductivity. With the best material in these studies being aluminum 7075, the experimental Electromagnetic Particle Injector (EPI) system has propelled

  7. Development of a High Fidelity System Analysis Code for Generation IV Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Vincent Mousseau; Haihua Zhao


    Traditional nuclear reactor system analysis codes such as RELAP and TRAC employ an operator split methodology. In this approach, each of the physics (fluid flow, heat conduction and neutron diffusion) is solved separately and the coupling terms are done explicitly. This approach limits accuracy (first order in time at best) and makes the codes slow in running since the explicit coupling imposes stability restrictions on the time step size. These codes have been extensively tested and validated for the existing LWRs. However, for GEN IV nuclear reactor designs which tend to have long lasting transients resulting from passive safety systems, the performance is questionable and modern high fidelity simulation tools will be required. The requirement for accurate predictability is the motivation for a large scale overhaul of all of the models and assumptions in transient nuclear reactor safety simulation software. At INL we have launched an effort with the long term goal of developing a high fidelity system analysis code that employs modern physical models, numerical methods, and computer science for transient safety analysis of GEN IV nuclear reactors. Modern parallel solution algorithms will be employed through utilizing the nonlinear solution software package PETSc developed by Argonne National Laboratory. The physical models to be developed will have physically realistic length scales and time scales. The solution algorithm will be based on the physics-based preconditioned Jacobian-free Newton-Krylov solution methods. In this approach all of the physical models are solved implicitly and simultaneously in a single nonlinear system. This includes the coolant flow, nonlinear heat conduction, neutron kinetics, and thermal radiation, etc. Including modern physical models and accurate space and time discretizations will allow the simulation capability to be second order accurate in space and in time. This paper presents the current status of the development efforts as

  8. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.


    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  9. Development concept for a small, split-core, heat-pipe-cooled nuclear reactor (United States)

    Lantz, E.; Breitwieser, R.; Niederauer, G. F.


    There have been two main deterrents to the development of semiportable nuclear reactors. One is the high development costs; the other is the inability to satisfy with assurance the questions of operational safety. This report shows how a split-core, heat-pipe cooled reactor could conceptually eliminate these deterrents, and examines and summarizes recent work on split-core, heat-pipe reactors. A concept for a small reactor that could be developed at a comparatively low cost is presented. The concept would extend the technology of subcritical radioisotope thermoelectric generators using 238 PuO2 to the evolution of critical space power reactors using 239 PuO2.

  10. Microcomputer-based equipment-control and data-acquisition system for fission-reactor reactivity-worth measurements

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, W.P.; Bucher, R.G.


    Material reactivity-worth measurements are one of the major classes of experiments conducted on the Zero Power research reactors (ZPR) at Argonne National Laboratory. These measurements require the monitoring of the position of a servo control element as a sample material is positioned at various locations in a critical reactor configuration. In order to guarantee operational reliability and increase experimental flexibility for these measurements, the obsolete hardware-based control unit has been replaced with a microcomputer based equipment control and data acquisition system. This system is based on an S-100 bus, dual floppy disk computer with custom built cards to interface with the experimental system. To measure reactivity worths, the system accurately positions samples in the reactor core and acquires data on the position of the servo control element. The data are then analyzed to determine statistical adequacy. The paper covers both the hardware and software aspects of the design.

  11. Development of MCATHAS system of coupled neutronics/thermal-hydraulics in supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    An, P.; Yao, D. [Science and Tech. on Reactor System Design Tech. Laboratory, Chengdu (China)


    The MCATHAS system of coupled neutronics/Thermal-hydraulics in supercritical water reactor is described, which considers the mutual influence between the obvious axial and radial evolution of material temperature, water density and the relative power distribution. This system can obtain the main neutronics and thermal parameters along with burn-up. MCATHAS system is parallel processing coupling. The MCNP code is used for neutronics analysis with the continuous cross section library at any temperature calculated by interpolation algorithm; The sub-channel code ATHAS is for thermal-hydraulics analysis and the ORIGEN Code for burn-up calculation. We validate the code with the assembly of HPLWR and analyze the assembly SCLWR- H. (author)

  12. Bifurcation in the Lengyel–Epstein system for the coupled reactors with diffusion

    Directory of Open Access Journals (Sweden)

    Shaban Aly


    Full Text Available The main goal of this paper is to continue the investigations of the important system of Fengqi et al. (2008. The occurrence of Turing and Hopf bifurcations in small homogeneous arrays of two coupled reactors via diffusion-linked mass transfer which described by a system of ordinary differential equations is considered. I study the conditions of the existence as well as stability properties of the equilibrium solutions and derive the precise conditions on the parameters to show that the Hopf bifurcation occurs. Analytically I show that a diffusion driven instability occurs at a certain critical value, when the system undergoes a Turing bifurcation, patterns emerge. The spatially homogeneous equilibrium loses its stability and two new spatially non-constant stable equilibria emerge which are asymptotically stable. Numerically, at a certain critical value of diffusion the periodic solution gets destabilized and two new spatially nonconstant periodic solutions arise by Turing bifurcation.

  13. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee


    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

  14. Development of Input/Output System for the Reactor Transient Analysis System (RETAS)

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jae Seung; Kang, Doo Hyuk; Cho, Yeon Sik [ENESYS, Daejeon (Korea, Republic of); Ahn, Seung Hoon; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)


    A Korea Institute of Nuclear Safety Reactor Transient Analysis System (KINS-RETAS) aims at providing a realistic prediction of core and RCS response to the potential or actual event scenarios in Korean nuclear power plants (NPPs). A thermal hydraulic system code MARS is a pivot code of the RETAS, and used to predict thermal hydraulic (TH) behaviors in the core and associated systems. MARS alone can be applied to many types of transients, but is sometimes coupled with the other codes developed for different objectives. Many tools have been developed to aid users in preparing input and displaying the transient information and output data. Output file and Graphical User Interfaces (GUI) that help prepare input decks, as seen in SNAP (Gitnick, 1998), VISA (K.D. Kim, 2007) and display aids include the eFAST (KINS, 2007). The tools listed above are graphical interfaces. The input deck builders allow the user to create a functional diagram of the plant, pictorially on the screen. The functional diagram, when annotated with control volume and junction numbers, is a nodalization diagram. Data required for an input deck is entered for volumes and junctions through a mouse-driven menu and pop-up dialog; after the information is complete, an input deck is generated. Display GUIs show data from MARS calculations, either during or after the transient. The RETAS requires the user to first generate a set of 'input', two dimensional pictures of the plant on which some of the data is displayed either numerically or with a color map. The RETAS can generate XY-plots of the data. Time histories of plant conditions can be seen via the plots or through the RETAS's replay mode. The user input was combined with design input from MARS developers and experts from both the GUI and ergonomics fields. A partial list of capabilities follows. - 3D display for neutronics. - Easier method (less user time and effort) to generate 'input' for the 3D displays. - Detailed view

  15. Requirement analysis and architecture of data communication system for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, K. I.; Kwon, H. J.; Park, J. H.; Park, H. Y.; Koo, I. S


    When digitalizing the Instrumentation and Control(I and C) systems in Nuclear Power Plants(NPP), a communication network is required for exchanging the digitalized data between I and C equipments in a NPP. A requirements analysis and an analysis of design elements and techniques are required for the design of a communication network. Through the requirements analysis of the code and regulation documents such as NUREG/CR-6082, section 7.9 of NUREG 0800 , IEEE Standard 7-4.3.2 and IEEE Standard 603, the extracted requirements can be used as a design basis and design concept for a detailed design of a communication network in the I and C system of an integral reactor. Design elements and techniques such as a physical topology, protocol transmission media and interconnection device should be considered for designing a communication network. Each design element and technique should be analyzed and evaluated as a portion of the integrated communication network design. In this report, the basic design requirements related to the design of communication network are investigated by using the code and regulation documents and an analysis of the design elements and techniques is performed. Based on these investigation and analysis, the overall architecture including the safety communication network and the non-safety communication network is proposed for an integral reactor.

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  17. The imprint of the Hawking effect in subcritical flows

    CERN Document Server

    Coutant, Antonin


    We study the propagation of low frequency shallow water waves on a one dimensional flow of varying depth. When taking into account dispersive effects, the linear propagation of long wavelength modes on uneven bottoms excites new solutions of the dispersion relation which possess a much shorter wavelength. The peculiarity is that one of these new solutions has a negative energy. When the flow becomes supercritical, this mode has been shown to be responsible for the (classical) analog of the Hawking effect. For subcritical flows, the production of this mode has been observed numerically and experimentally, but the precise physics governing the scattering remained unclear. In this work, we provide an analytic treatment of this effect in subcritical flows. We analyze the scattering of low frequency waves using a new perturbative series, derived from a generalization of the Bremmer series. We show that the production of short wavelength modes is governed by a complex value of the position: a complex turning point....

  18. A microfluidic sub-critical water extraction instrument (United States)

    Sherrit, Stewart; Noell, Aaron C.; Fisher, Anita; Lee, Mike C.; Takano, Nobuyuki; Bao, Xiaoqi; Kutzer, Thomas C.; Grunthaner, Frank


    This article discusses a microfluidic subcritical water extraction (SCWE) chip for autonomous extraction of amino acids from astrobiologically interesting samples. The microfluidic instrument is composed of three major components. These include a mixing chamber where the soil sample is mixed and agitated with the solvent (water), a subcritical water extraction chamber where the sample is sealed with a freeze valve at the chip inlet after a vapor bubble is injected into the inlet channels to ensure the pressure in the chip is in equilibrium with the vapor pressure and the slurry is then heated to ≤200 °C in the SCWE chamber, and a filter or settling chamber where the slurry is pumped to after extraction. The extraction yield of the microfluidic SCWE chip process ranged from 50% compared to acid hydrolysis and 80%-100% compared to a benchtop microwave SCWE for low biomass samples.

  19. Software development methodology for computer based I&C systems of prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Manimaran, M., E-mail:; Shanmugam, A.; Parimalam, P.; Murali, N.; Satya Murty, S.A.V.


    Highlights: • Software development methodology adopted for computer based I&C systems of PFBR is detailed. • Constraints imposed as part of software requirements and coding phase are elaborated. • Compliance to safety and security requirements are described. • Usage of CASE (Computer Aided Software Engineering) tools during software design, analysis and testing phase are explained. - Abstract: Prototype Fast Breeder Reactor (PFBR) is sodium cooled reactor which is in the advanced stage of construction in Kalpakkam, India. Versa Module Europa bus based Real Time Computer (RTC) systems are deployed for Instrumentation & Control of PFBR. RTC systems have to perform safety functions within the stipulated time which calls for highly dependable software. Hence, well defined software development methodology is adopted for RTC systems starting from the requirement capture phase till the final validation of the software product. V-model is used for software development. IEC 60880 standard and AERB SG D-25 guideline are followed at each phase of software development. Requirements documents and design documents are prepared as per IEEE standards. Defensive programming strategies are followed for software development using C language. Verification and validation (V&V) of documents and software are carried out at each phase by independent V&V committee. Computer aided software engineering tools are used for software modelling, checking for MISRA C compliance and to carry out static and dynamic analysis. Various software metrics such as cyclomatic complexity, nesting depth and comment to code are checked. Test cases are generated using equivalence class partitioning, boundary value analysis and cause and effect graphing techniques. System integration testing is carried out wherein functional and performance requirements of the system are monitored.

  20. Comprehensive Prediction of Thermosyphon Characteristics in Reactor Passive Cooling System Simulation Loop FASSIP-01

    Directory of Open Access Journals (Sweden)

    H. Tjahjono


    Full Text Available Passive cooling mechanism for a nuclear reactor has been proven to be very important since the Fukushima Daiichi Reactor accident that was caused by active cooling system malfunction due to total loss of electrical power source. In the Center for Nuclear Reactor Technology and Safety of BATAN, the cooling mechanism was studied by using a natural circulation test loop named FASSIP-01 that applied thermosyphon mechan