WorldWideScience

Sample records for water reactor concepts

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

    CERN Document Server

    Kessler, Günter; Schlüter, Franz-Hermann

    2014-01-01

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

  2. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  3. Integral Fast Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  4. Study on the Use of Hydride Fuel in High-Performance Light Water Reactor Concept

    Directory of Open Access Journals (Sweden)

    Haileyesus Tsige-Tamirat

    2015-01-01

    Full Text Available Hydride fuels have features which could make their use attractive in future advanced power reactors. The potential benefit of use of hydride fuel in HPLWR without introducing significant modification in the current core design concept of the high-performance light water reactor (HPLWR has been evaluated. Neutronics and thermal hydraulic analyses were performed for a single assembly model of HPLWR with oxide and hydride fuels. The hydride assembly shows higher moderation with softer neutron spectrum and slightly more uniform axial power distribution. It achieves a cycle length of 18 months with sufficient excess reactivity. At Beginning of Cycle the fuel temperature coefficient of the hydride assembly is higher whereas the moderator and void coefficients are lower. The thermal hydraulic results show that the achievable fuel temperature in the hydride assembly is well below the design limits. The potential benefits of the use of hydride fuel in the current design of the HPLWR with the achieved improvements in the core neutronics characteristics are not sufficient to justify the replacement of the oxide fuel. Therefore for a final evaluation of the use of hydride fuels in HPLWR concepts additional studies which include modification of subassembly and core layout designs are required.

  5. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  6. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept...reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used

  7. A concept of JAERI passive safety light water reactor system (JPSR)

    Energy Technology Data Exchange (ETDEWEB)

    Murao, Y.; Araya, F.; Iwamura, T. [Japan Atomic Energy Research Institute, Tokai-mura (Japan)

    1995-09-01

    The Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor system concept, JPSR, which was developed for reducing manpower in operation and maintenance and influence of human errors on reactor safety. In the concept the system was extremely simplified. The inherent matching nature of core generation and heat removal rate within a small volume change of the primary coolant is introduced by eliminating chemical shim and adopting in-vessel control rod drive mechanism units, a low power density core and once-through steam generators. In order to simplify the system, a large pressurizer, canned pumps, passive engineered-safety-features-system (residual heat removal system and coolant injection system) are adopted and the total system can be significantly simplified. The residual heat removal system is completely passively actuated in non-LOCAs and is also used for depressurization of the primary coolant system to actuate accumulators in small break LOCAs and reactor shutdown cooling system in normal operation. All of systems for nuclear steam supply system are built in the containment except for the air coolers as a the final heat sink of the passive residual heat removal system. Accordingly the reliability of the safety system and the normal operation system is improved, since most of residual heat removal system is always working and a heat sink for normal operation system is {open_quotes}safety class{close_quotes}. In the passive coolant injection system, depressurization of the primary cooling system by residual heat removal system initiates injection from accumulators designed for the MS-600 in medium pressure and initiates injection from the gravity driven coolant injection pool at low pressure. Analysis with RETRAN-02/MOD3 code demonstrated the capability of passive load-following, self-power-controllability, cooling and depressurization.

  8. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  9. Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    R. Sonat Sen; Michael A. Pope; Abderrafi M. Ougouag; Kemal O. Pasamehmetoglu

    2012-04-01

    The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities [1]. Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention [2]. The Deep Burn project [3] currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

  10. New reactor concepts; Nieuwe rectorconcepten - nouveaux reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost.

  11. Evolution of the tandem mirror reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, G.A.; Logan, B.G.

    1982-03-09

    We discuss the evolution of the tandem mirror reactor concept from the original conceptual reactor design (1977) through the first application of the thermal barrier concept to a reactor design (1979) to the beginning of the Mirror Advanced Reactor Study (1982).

  12. Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Li-Puma, Antonella, E-mail: antonella.lipuma@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Richou, Marianne; Magaud, Philippe; Missirlian, Marc [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Visca, Eliseo [Associazione EURATOM-ENEA sulla Fusione, C.R. Frascati, IT-00044 Frascati (Italy); Ridolfini, Vincenzo Pericoli [EFDA-CSU Garching, PPPT department, D-85748 Garching bei München (Germany)

    2013-10-15

    In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies. For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed. Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R and D activities are reported. This work has been carried out in the frame of EFDA PPPT Work Programme activities.

  13. Heavy Water Reactor; Reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Yu, St.; HOpwood, J.; Meneley, D. [Energie Atomique du Canada (Canada)

    2000-04-01

    This document deals with the Heavy Water Reactor (HWR) technology and especially the Candu (Canada Deuterium Uranium) reactor. This reactors type offers many advantages that promote them for the future. General concepts, a description of the Candu nuclear power plants, the safety systems, the fuel cycle and economical and environmental aspects are included. (A.L.B.)

  14. Light water reactor program

    Energy Technology Data Exchange (ETDEWEB)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  15. Developments and Tendencies in Fission Reactor Concepts

    Science.gov (United States)

    Adamov, E. O.; Fuji-Ie, Y.

    This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC

  16. Design of a boiling water reactor core based on an integrated blanket-seed thorium-uranium concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Mexico, D.F. (Mexico); Francois, Juan Luis [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico)]. E-mail: jlfl@fi-b.unam.mx; Martin-del-Campo, Cecilia [Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana, Avenida San Rafael Atlixco 186, Col. Vicentina, 09340 Mexico, D.F. (Mexico)

    2005-04-15

    This paper is concerned with the design of a boiling water reactor (BWR) equilibrium core using thorium as a nuclear material in an integrated blanket-seed (BS) assembly. The integrated BS concept comes from the fact that the blanket and the seed rods are located in the same assembly, and are burned out in a once-through cycle. The idea behind the lattice design is to use the thorium conversion capability in a BWR spectrum, taking advantage of the {sup 233}U build-up. A core design was developed to achieve an equilibrium cycle of 365 effective full power days in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main operating parameters, like power, linear heat generation rate and void distributions were obtained as well as the shutdown margin. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The shutdown margin design criterion was fulfilled by addition of a burnable poison region in the fuel assembly.

  17. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

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

  18. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  19. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  20. ASME Material Challenges for Advanced Reactor Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  1. Concept for LEU Burst Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Steven Karl [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kimpland, Robert Herbert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    Design and performance of a proposed LEU burst reactor are sketched. Salient conclusions reached are the following: size would be ~1,500 kg or greater, depending on the size of the central cavity; internal stresses during burst require split rings for relief; the reactor would likely require multiple control and safety rods for fine control; the energy spectrum would be comparable to that of HEU machines; and burst yields and steady-state power levels will be significantly greater in an LEU reactor.

  2. Study on water cooled high conversion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    As a part of study on advanced reactors for the future, conceptual design of high conversion water cooled reactors is being studied, aiming at the contribution to nuclear fuel cycle by the LWR technology, since the utilization of LWRs will extend over a long period of time . We are studying on the reactor core concepts for BWR and PWR reactor systems. As for BWR system, three types of reactor cores are investigating for three different design goals; long operation period, high conversion ratio and high applicability for the existing BWR system. In all the cases, we have obtained a fair prospect of a large core concept with a capacity of 1,000 MWe class having negative void reactivity coefficient. This study is a part of JAERI-JAPCO (Japan Atomic Power Company) cooperative studies. Various kinds of conceptual designs will be created until the end of FY 1999. The designs will be checked and reviewed at that time, then experimental studies on the realization of the concepts will start with further design works from FY 2000. (author)

  3. New concepts for shaftless recycle reactors

    Energy Technology Data Exchange (ETDEWEB)

    Berty, J.M.; Berty, I.J.

    1987-01-01

    Berty Reaction Engineers, Ltd. (BREL) is developing two new laboratory recycle reactors, the ROTOBERTY and the TURBOBERTY. These new reactors are basically improved versions of the original Berty reactor. To understand why the reactors have the features that they do, it is first necessary to briefly review laboratory reactors in general and specifically the original Berty reactor.

  4. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  5. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  6. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  7. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  8. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. 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. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  9. PX–An Innovative Safety Concept for an Unmanned Reactor

    Directory of Open Access Journals (Sweden)

    Sung-Jae Yi

    2016-02-01

    Full Text Available An innovative safety concept for a light water reactor has been developed at the Korea Atomic Energy Research Institute. It is a unique concept that adopts both a fast heat transfer mechanism for a small containment and a changing mechanism of the cooling geometry to take advantage of the potential, thermal, and dynamic energies of the cold water in the containment. It can bring about rapid cooling of the containment and long-term cooling of the decay heat. By virtue of this innovative concept, nuclear fuel damage events can be prevented. The ultimate heat transfer mechanism contributes to minimization of the heat exchanger size and containment volume. A small containment can ensure the underground construction, which can use river or seawater as an ultimate heat sink. The changing mechanism of the cooling geometry simplifies several safety systems and unifies diverse functions. Simplicity of the present safety system does not require any operator actions during events or accidents. Therefore, the unique safety concept of PX can realize both economic competitiveness and inherent safety.

  10. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  11. Utilization of plutonium in reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

    Japan's nuclear policy decides not to have excess plutonium. Upon assuming the future situation of the delay of FBR introduction, the JAERI performs the feasibility study of several types of the reduced-moderation water reactors (RMWRs). As the RMWRs have higher conversion ratio than LWRs, they are expected to enable multi-cycle utilization of plutonium, high burnup and long cycle operation, and enhancement of uranium resource utilization. While the full MOX LWRs are being developed, from viewpoint of suppressing the accumulation of plutonium, the RMWRs are thought to be more suitable. As plutonium inventory is larger in the RMWRs than in the full MOX LWRs, also from viewpoint of non-proliferation of nuclear materials, the RMWRs are thought to be more suitable. The current feasibility study will be performed until 2010 to confirm the position, to construct the reactor concept, and to demonstrate the feasibility on reactor physics and on thermal hydraulics. The present candidate reactor types of the study are three BWR types, heavy water cooled PWR type and light water cooled PWR type. Hereafter comprehensive evaluation from viewpoint of problems on fuel cycle, economy, continuity with conventional LWR technologies will be performed to extract the most suitable concept to satisfy the social needs and to construct the fundamental reactor concept to concentrate R and D effort. (K. Tsuchihashi)

  12. Advanced Nuclear Reactor Concepts for China

    Energy Technology Data Exchange (ETDEWEB)

    Knoche, D.; Sassen, F.; Tietsch, W. [Westinghouse Electric Germany, Postfach 10 05 63, 68140 Mannheim (Germany); Yujie, Dong; Li, Cao [INET, Tsinghua University, 100084 Beijing (China)

    2008-07-01

    China is one of the fastest growing economies in the world. With 1.3 billion people China also has the largest population worldwide. The growing economy, the migration of people from rural areas to cities and the augmentation in living standard will drive the energy demand of China in the coming decades. At present the installed electrical power is about 500 GW. In the years 2004 and 2005 the added electrical capacity was around 60 GW per year. Chinas primary energy demand is covered mainly by the use of coal. Coal also will remain the main energy source in the coming decades in China. Nevertheless taking into account more and more environmental aspects and the goal to reduce dependencies on energy imports a better energy mix strategy is planed to change including at an increasing level the renewable and nuclear option. Present the nuclear park is characterised by a large variety of different types of reactors. With the AP-1000, EPR and the gas-cooled High Temperature Reactor (HTR) the spectrum of different reactor types will be further enlarged. (authors)

  13. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    OpenAIRE

    Matjaž Leskovar; Mitja Uršič

    2016-01-01

    A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In ...

  14. Novel Photocatalytic Reactor Development for Removal of Hydrocarbons from Water

    Directory of Open Access Journals (Sweden)

    Morgan Adams

    2008-01-01

    Full Text Available Hydrocarbons contamination of the marine environment generated by the offshore oil and gas industry is generated from a number of sources including oil contaminated drill cuttings and produced waters. The removal of hydrocarbons from both these sources is one of the most significant challenges facing this sector as it moves towards zero emissions. The application of a number of techniques which have been used to successfully destroy hydrocarbons in produced water and waste water effluents has previously been reported. This paper reports the application of semiconductor photocatalysis as a final polishing step for the removal of hydrocarbons from two waste effluent sources. Two reactor concepts were considered: a simple flat plate immobilised film unit, and a new rotating drum photocatalytic reactor. Both units proved to be effective in removing residual hydrocarbons from the effluent with the drum reactor reducing the hydrocarbon content by 90% under 10 minutes.

  15. Advanced Burner Reactor 1000MWth Reference Concept

    Energy Technology Data Exchange (ETDEWEB)

    Cahalan, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fanning, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Kim, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Kellogg, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Krajtl, L. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Momozaki, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Park, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Reed, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Salev, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Seidensticker, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Tang, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Tzanos, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Wei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Chikazawa, Y. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2007-09-30

    The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence, to validate the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat.

  16. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

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

    Energy Technology Data Exchange (ETDEWEB)

    Nieto, M.; Miley, G.H. [Illinois Univ., Fusion Studies Lab., Dept. of Nuclear, Plasma, and Radiological Engineering, Urbana, IL (United States)

    2001-07-01

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

  18. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  19. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  20. Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

  1. Ex-vessel Steam Explosion Analysis for Pressurized Water Reactor and Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Matjaž Leskovar

    2016-02-01

    Full Text Available A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel–coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  3. Concept design on RH maintenance of CFETR Tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Wu, Songtao; Wan, Yuanxi; Li, Jiangang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Ye, Minyou [University of Science and Technology of China, Hefei (China); Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2014-10-15

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.

  4. Evaluation of the breed/burn fast reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Atefi, B.; Driscoll, M.J.; Lanning, D.D.

    1979-12-01

    A core design concept and fuel management strategy, designated breed/burn, has been evaluated for heterogeneous fast breeder reactors. In this concept internal blanket assemblies after fissile material is bred in over several incore cycles, are shuffled into a moderated radial blanket and/or central island. The most promising materials combination identified used thorium in the internal blankets (due to the superior performance of epithermal Th-U233 systems) and zirconium hydride (ZrH/sub 16/) as the moderator (because of the compact assembly and core designs it permitted).

  5. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  6. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T.; Grunwald, G.

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  7. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma (Computational Engineering Analysis, Albuquerque, NM); Al Rashdan, Ahmad (Texas A& M University, College Station, TX); Tsvetkov, Pavel Valeryevich (Texas A& M University, College Station, TX); Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  8. Computational Fluid Dynamics Analysis of Canadian Supercritical Water Reactor (SCWR)

    Science.gov (United States)

    Movassat, Mohammad; Bailey, Joanne; Yetisir, Metin

    2015-11-01

    A Computational Fluid Dynamics (CFD) simulation was performed on the proposed design for the Canadian SuperCritical Water Reactor (SCWR). The proposed Canadian SCWR is a 1200 MW(e) supercritical light-water cooled nuclear reactor with pressurized fuel channels. The reactor concept uses an inlet plenum that all fuel channels are attached to and an outlet header nested inside the inlet plenum. The coolant enters the inlet plenum at 350 C and exits the outlet header at 625 C. The operating pressure is approximately 26 MPa. The high pressure and high temperature outlet conditions result in a higher electric conversion efficiency as compared to existing light water reactors. In this work, CFD simulations were performed to model fluid flow and heat transfer in the inlet plenum, outlet header, and various parts of the fuel assembly. The ANSYS Fluent solver was used for simulations. Results showed that mass flow rate distribution in fuel channels varies radially and the inner channels achieve higher outlet temperatures. At the outlet header, zones with rotational flow were formed as the fluid from 336 fuel channels merged. Results also suggested that insulation of the outlet header should be considered to reduce the thermal stresses caused by the large temperature gradients.

  9. Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

    Directory of Open Access Journals (Sweden)

    Peel Ross

    2016-01-01

    Full Text Available This paper discusses concepts for three-component fuel bundles containing plutonium, uranium and thorium for use in pressurised heavy water reactors, and cases for and against implementation of such a nuclear energy system in the United Kingdom. Heavy water reactors are used extensively in Canada, and are deploying within India and China, whilst the UK is considering the use of heavy water reactors to manage its plutonium inventory of 140 tonnes. The UK heavy water reactor proposal uses a mixed oxide (MOX fuel of plutonium in depleted uranium, within the enhanced CANDU-6 (EC-6 reactor. This work proposes an alternative heterogeneous fuel concept based on the same reactor and CANFLEX fuel bundle, with eight large-diameter fuel elements loaded with natural thorium oxide and 35 small-diameter fuel elements loaded with a MOX of plutonium and reprocessed uranium stocks from UK MAGNOX and AGR reactors. Indicative neutronic calculations suggest that such a fuel would be neutronically feasible. A similar MOX may alternatively be fabricated from reprocessed <5% enriched light water reactor fuel, such as the fuel of the AREVA EPR reactor, to consume newly produced plutonium from reprocessing, similar to the DUPIC (direct use of PWR fuel in CANDU process.

  10. The material concept in German light water reactors. Contribution to plant safety economic efficiency and failure provision; Das Werkstoffkonzept in deutschen Leichtwasserreaktoren. Beitrag zur Anlagensicherheit, Wirtschaftlichkeit und Schadensvorsorge

    Energy Technology Data Exchange (ETDEWEB)

    Ilg, Ulf [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Philippsburg; Koenig, Guenter [EnBW Kernkraft GmbH (Germany). Kernkraftwerk Neckarwestheim; Erve, Manfred [AREVA NP GmbH, Erlangen (Germany)

    2008-07-01

    In the design and construction stage of nuclear power plants relevant decisions may affect the service life of a component, and thus influence safety and availability of the plant. The German ''basic safety concept'' has an important effect on the quality of the BOL (begin of life) status. Materials selection and qualification are of significant importance for the component lifetime and the profitability of the plant. Examples for the implementation of this concept are demonstrated for the steam generator tubing material Incoloy 800, the inside-plated ferritic compound tubes as control rod drive mechanism nozzle through the RPV head of BWR plants that are not susceptible for corrosion enhanced cracking that was observed for Inconel 600 tubing. A fundamental failure analysis of crack formation in Ti stabilized austenitic pipes of BWR plants found since 1993 were definitely identified as intergranular stress corrosion caused by a local sensitization of the welding process induced overheated structured in the heat affected zone. This allowed target-oriented mitigation measures. The safety culture implemented in German nuclear plants in connection with the break preclusion or integrity concept, respectively, including a continuous actualization with respect to the state-of-the art are the technical prerequisites for damage precaution and possible life time extension.

  11. New reactor concepts. An analysis of the actual research status; Neue Reaktorkonzepte. Eine Analyse des aktuellen Forschungsstands

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias

    2017-04-15

    The report on new reactor concepts covers the following issues: characterization and survey of new reactor concepts; evaluation criteria: safety, resources for fuel supply, waste problems, economy and proliferation; comprehensive relevant aspects: thorium as alternative resource, partitioning and transmutation; actual developments and preliminary experiences for fast breeding reactor (FBR), high-temperature reactor (HTR), molten salt reactor (MSR), small modular reactor (SMR).

  12. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    Science.gov (United States)

    Treshow, M.

    1963-04-30

    This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

  13. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-08-08

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

  14. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  15. Pellet bed reactor concept for nuclear electric propulsion

    Science.gov (United States)

    El-Genk, Mohamed S.; Morley, Nicholas J.; Juhasz, Albert

    1993-01-01

    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

  16. Multi-Applications Small Light Water Reactor - NERI Final Report

    Energy Technology Data Exchange (ETDEWEB)

    S. Michale Modro; James E. Fisher; Kevan D. Weaver; Jose N. Reyes, Jr.; John T. Groome; Pierre Babka; Thomas M. Carlson

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle.

  17. Advanced light water reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    Giedraityte, Zivile [Helsinki University of Technology, Otaranta 8D-84, 02150 Espoo (Finland)

    2008-07-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  18. Silicon carbide composite for light water reactor fuel assembly applications

    Science.gov (United States)

    Yueh, Ken; Terrani, Kurt A.

    2014-05-01

    The feasibility of using SiCf-SiCm composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  19. Silicon carbide composite for light water reactor fuel assembly applications

    Energy Technology Data Exchange (ETDEWEB)

    Yueh, Ken, E-mail: kyueh@epri.com [Fuel Reliability Program, EPRI, 1300 West WT Harris Blvd, Charlotte, NC 28262 (United States); Terrani, Kurt A., E-mail: terranika@ornl.gov [Fusion and Materials for Nuclear Systems Division, Oak Ridge National Laboratory, 1 Bethel Valley Rd. MS 6093, Oak Ridge, TN 37831 (United States)

    2014-05-01

    The feasibility of using SiC{sub f}–SiC{sub m} composites in light water reactor (LWR) fuel designs was evaluated. The evaluation was motivated by the desire to improve fuel performance under normal and accident conditions. The Fukushima accident once again highlighted the need for improved fuel materials that can maintain fuel integrity to higher temperatures for longer periods of time. The review identified many benefits as well as issues in using the material. Issues perceived as presenting the biggest challenges to the concept were identified to be flux gradient induced differential volumetric swelling, fragmentation and thermal shock resistance. The oxidation of silicon and its release into the coolant as silica has been identified as an issue because existing plant systems have limited ability for its removal. Detailed evaluation using available literature data and testing as part of this evaluation effort have eliminated most of the major concerns. The evaluation identified Boiling Water Reactor (BWR) channel, BWR fuel water tube, and Pressurized Water Reactor (PWR) guide tube as feasible applications for SiC composite. A program has been initiated to resolve some of the remaining issues and to generate physical property data to support the design of commercial fuel components.

  20. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  1. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  2. Fast reactor core concepts to improve transmutation efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Fujimura, Koji; Kawashima, Katsuyuki [Hitachi Research Laboratory, Hitachi, Ltd., 7-1-1, Omika-cho, Hitachi-shi, Ibaraki, 319-1221 Japan (Japan); Itooka, Satoshi [Hitachi-GE Nuclear Energy, Ltd., 3-1-1, Saiwai-cho, Hitachi-shi, Ibaraki, 317-0073 Japan (Japan)

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  3. Concept of magnet systems for LHD-type reactor

    Science.gov (United States)

    Imagawa, S.; Takahata, K.; Tamura, H.; Yanagi, N.; Mito, T.; Obana, T.; Sagara, A.

    2009-07-01

    Heliotron reactors have attractive features for fusion power plants such as having no need for current drive and a wide space between the helical coils for the maintenance of in-vessel components. Their main disadvantage was considered to be the necessarily large size of their magnet systems. According to the recent reactor studies based on the experimental results in the Large Helical Device, a major radius of plasma of 14-17 m with a central toroidal field of 6-4 T is needed to attain the self-ignition condition with a blanket space thicker than 1.1 m. The stored magnetic energy is estimated at 120-140 GJ. Although both the major radius and the magnetic energy are about three times as large as ITER, the maximum magnetic field and mechanical stress are comparable. In the preliminary structural analysis, the maximum stress intensity including the peak stress is less than the 1000 MPa that is allowed for strengthened stainless steel. Although the length of the helical coil is more than 150 m, that is about five times as long as the ITER TF coil, cable-in-conduit conductors can be adopted with a parallel winding method of five-in-hand. The concept of the parallel winding is proposed. Consequently, the magnet systems for helical reactors can be realized with a small extension of the ITER technology.

  4. Preliminary design concept of a subcritical reactor using available resources

    Energy Technology Data Exchange (ETDEWEB)

    Churnetski, E.L. [Oak Ridge Y-12 Plant, TN (United States); Hoyny, V.; Chaudhuri, B.R.; Taprantzis, A.; Yavas, A. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    1993-12-31

    During the Fall 1993 semester, a project was initiated within the Nuclear Engineering Department of the University of Tennessee with the objective of developing a design for a subcritical reactor with maximized multiplication factor using materials currently available. Such a device, if constructed, would serve as a teaching tool for the Department of Nuclear Engineering. Design work was conducted as a large number of computer calculations, with trial pile configurations based on fundamental nuclear engineering principles, in an effort to maximize multiplication factor through fuel element geometry, moderator type, fissile/moderator ratio, and reflector character. The principal objective of the design group for the early phase of this project was to present several possible ``baseline`` reactor designs and identify directions for improvements. For the sake of calculational ease, the cores analyzes to date have been of nearly cubic shape. The SCALE CSAS25 software which runs KENO.Va, a Monte Carlo code, was used for all neutronics calculations. The baseline reactors resulting from work to date are cuboidal in shape and graphite reflected. Two types of fuel element geometries are proposed, a typical triangular pitch rod lattice and an arrangement of discrete fuel slugs placed in a lattice corresponding to body centered cubic packing. The latter arrangement provides slightly higher multiplication factors than the former. Calculations were performed for both graphite and heavy water moderation with heavy water moderation producing considerably higher multiplication factors, as expected. In general, the maximum k{sub eff} for the reactors are in the range of 0.5 to 0.9, well subcritical, except in the cases of the extreme possible values of fuel assay where critical configurations are possible. In these cases, designs with reduced fuel loading are recommended to assure a subcritical multiplication factor.

  5. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  6. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  7. First wall and blanket concepts for experimental fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Biggio, M.; Cardella, A.; Daenner, W.; Farfaletti-Casali, F.; Ponti, C.; Rieger, M.; Vieider, G.

    1985-07-01

    The paper describes the progress of the studies on first wall and liquid breeder blankets for tritium production in the Next European Torus (NET). Two concepts of first wall/blanket segments are described, using 17Li83Pb as breeder and water as coolant. In both concepts the first wall is integrated in a steel box enveloping the breeder units which are cylindrical vessels with an inside heat transfer system. The thermomechanical and neutronics features of the two concepts are evaluated. Finally, the questions related to tritium permeation into coolant and tritium recovery from breeder are discussed on the basis of the analysis in progress in Europe.

  8. The use of PRA in the development of ALWR (advanced light water reactor) design requirements. [Advanced Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Summitt, R.L. (Safety and Reliability Optimization Services, Inc., Knoxville, TN (USA)); Additon, S.L. (TENERA, L.P., Bethesda, MD (USA)); Pasedag, W.F. (USDOE, Washington, DC (USA))

    1989-01-01

    The current hiatus in nuclear power plant orders provides an opportunity for the development of advanced light water reactor (ALWR) design concepts and regulatory requirements which incorporate the insights gained from the application of the probabilistic risk assessment. The US Department of Energy is assisting the Electric Power Research Institute (EPRI) in the incorporation of PRA insights into the specification of the utility requirements, and reactor vendors in support of the conceptual design of safety systems, for such advanced plants. This paper reviews the applications of PRA methods in this development of specifications for, and the design of simplified, rugged ALWRs with a significantly improved risk profile. Specific examples of the impact of utilizing published PRA insights, construction and use of functional PRA models, and feedback of PRA experience into the specification of the key assumptions and groundrules for ALWR PRAs are presented. 13 refs., 3 tabs.

  9. REACTOR - a Concept for establishing a System-of-Systems

    Science.gov (United States)

    Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

    2014-05-01

    well suited to establish brokers, which mediate metadata and semantic information about the resources of all involved systems. This concept has been developed within the project Collaborative, Complex, and Critical Decision-Support in Evolving Crises (TRIDEC) on the basis of semantic registries describing all facets of events and services utilisable for crisis management systems. The implementation utilises an operative infrastructure including an Enterprise Service Bus (ESB), adapters to proprietary sensor systems, a workflow engine, and a broker-based MOM. It also applies current technologies like actor-based frameworks for highly concurrent, distributed, and fault tolerant event-driven applications. Therefore REACTOR implementations are well suited to be hosted in a cloud that provides Infrastructure as a Service (IaaS). To provide low entry barriers for legacy and future systems, REACTOR adapts the principles of Design by Contract (DbC) as well as standardised and common information models like the Sensor Web Enablement (SWE) or the JavaScript Object Notation for geographic features (GeoJSON). REACTOR has been applied exemplarily within two different scenarios, Natural Crisis Management and Industrial Subsurface Development.

  10. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

    2001-07-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  11. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  12. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  13. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  14. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  15. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  16. Environmentally assisted cracking in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  17. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    Science.gov (United States)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  18. Concept Design of Supercritical Water Cooled Reactor Core with Double-Row-Rod Assemblies%双排棒组件超临界水堆堆芯方案设计

    Institute of Scientific and Technical Information of China (English)

    杨珏; 张勇; 赵传奇; 单建强; 王飞; 曹良志

    2012-01-01

    结合国际上多种超临界水堆堆芯设计方案的优点,提出了一种新的压力容器式低泄漏堆芯设计方案,其特点是,堆芯中采用了双排棒正方形闭式燃料组件和三区低泄漏换料.双排棒燃料组件由两排燃料棒包围一个慢化剂水棒构成,可以使得慢化均匀;三区低泄漏换料可以大大延长堆芯寿期,降低压力容器快中子注量.通过堆芯三维物理热工耦合计算发现,该方案寿期内的最大包壳温度(MCST)为684℃,堆芯寿期为300个有效满功率天,且功率分布平坦.在此基础上,对所有组件进行了更为保守的子通道热工水力计算,得出MCST为685.3℃,进一步表明所提堆芯设计方案在物理热工方面是可行的.%A new pressure-vessel type supercritical water cooled reactor (SCWR) core concept was proposed by combining merits of several SCWR core designs in the world. This core design employs a new type of closed assembly with double-row fuel rods in square geometry and a three-batch low-leakage refueling scheme. The assembly consists of two rows of fuel rods and a moderator rod, which causes the moderation more uniform. The three-batch low leakage refueling scheme obviously increases the cycle length and reduces the neutron fluence on the pressure vessel. Three-dimensional neutronics/thermohydraulics coupling calculation shows that the maximum cladding surface temperature (MCST) is 684 ℃, the cycle life is 300 effective full power days and the power distribution is flat. Then the more conservative sub-channel analysis was performed for all fuel assemblies. The MCST was evaluated to be 685. 3 ℃, showing that the core design is feasible.

  19. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  20. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

  1. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  2. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  3. Computer simulation of the NASA water vapor electrolysis reactor

    Science.gov (United States)

    Bloom, A. M.

    1974-01-01

    The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

  4. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  5. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.

  6. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-11-01

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

  8. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

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

  9. Advanced Concepts for Pressure-Channel Reactors: Modularity, Performance and Safety

    Science.gov (United States)

    Duffey, Romney B.; Pioro, Igor L.; Kuran, Sermet

    Based on an analysis of the development of advanced concepts for pressure-tube reactor technology, we adapt and adopt the pressure-tube reactor advantage of modularity, so that the subdivided core has the potential for optimization of the core, safety, fuel cycle and thermal performance independently, while retaining passive safety features. In addition, by adopting supercritical water-cooling, the logical developments from existing supercritical turbine technology and “steam” systems can be utilized. Supercritical and ultra-supercritical boilers and turbines have been operating for some time in coal-fired power plants. Using coolant outlet temperatures of about 625°C achieves operating plant thermal efficiencies in the order of 45-48%, using a direct turbine cycle. In addition, by using reheat channels, the plant has the potential to produce low-cost process heat, in amounts that are customer and market dependent. The use of reheat systems further increases the overall thermal efficiency to 55% and beyond. With the flexibility of a range of plant sizes suitable for both small (400 MWe) and large (1400 MWe) electric grids, and the ability for co-generation of electric power, process heat, and hydrogen, the concept is competitive. The choice of core power, reheat channel number and exit temperature are all set by customer and materials requirements. The pressure channel is a key technology that is needed to make use of supercritical water (SCW) in CANDU®1 reactors feasible. By optimizing the fuel bundle and fuel channel, convection and conduction assure heat removal using passive-moderator cooling. Potential for severe core damage can be almost eliminated, even without the necessity of activating the emergency-cooling systems. The small size of containment structure lends itself to a small footprint, impacts economics and building techniques. Design features related to Canadian concepts are discussed in this paper. The main conclusion is that development of

  10. Environmentally assisted cracking in light water reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

    2007-11-06

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2002. Topics that have been investigated include: (a) environmental effects on fatigue crack initiation in carbon and low-alloy steels and austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs in BWRs, (c) evaluation of causes and mechanisms of irradiation-assisted cracking of austenitic SS in PWRs, and (d) cracking in Ni-alloys and welds. A critical review of the ASME Code fatigue design margins and an assessment of the conservation in the current choice of design margins are presented. The existing fatigue {var_epsilon}-N data have been evaluated to define the effects of key material, loading, and environmental parameters on the fatigue lives of carbon and low-alloy steels and austenitic SSs. Experimental data are presented on the effects of surface roughness on fatigue crack initiation in these materials in air and LWR environments. Crack growth tests were performed in BWR environments on SSs irradiated to 0.9 and 2.0 x 10{sup 21} n x cm{sup -2}. The crack growth rates (CGRs) of the irradiated steels are a factor of {approx}5 higher than the disposition curve proposed in NUREG-0313 for thermally sensitized materials. The CGRs decreased by an order of magnitude in low-dissolved oxygen (DO) environments. Slow-strain-rate tensile (SSRT) tests were conducted in high-purity 289 C water on steels irradiated to {approx}3 dpa. The bulk S content correlated well with the susceptibility to intergranular SCC in 289 C water. The IASCC susceptibility of SSs that contain >0.003 wt. % S increased drastically. bend tests in inert environments at 23 C were conducted on broken pieces of SSRT specimens and on unirradiated specimens of the same materials after hydrogen charging. The results of the tests and a review of other data in the literature

  11. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

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

  12. Shielding designs for pressurized water reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Forestier, J.; Vergnaud, T.

    1986-07-01

    The efforts made by Electricite de France to reduce exposure from the two-component neutron-gamma radiation fields inside the pressurized water reactor (PWR) building are described. Most of the attention had been focused on the problem of neutron exposure relative to the problem of achieving a highly efficient confinement within the reactor cavity and the state of the art of personnel neutron dosimetry. A description of the general neutron calculation scheme that links the characteristics of the neutron fields escaping from the reactor vessel to the dose equivalent rate cartographies inside the reactor building is provided.

  13. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  14. SABR fusion-fission hybrid transmutation reactor design concept

    Science.gov (United States)

    Stacey, Weston

    2009-11-01

    A conceptual design has been developed for a sub-critical advanced burner reactor (SABR) consisting of i) a sodium cooled fast reactor fueled with the transuranics (TRU) from spent nuclear fuel, and ii) a D-T tokamak fusion neutron source based on ITER physics and technology. Subcritical operation enables more efficient transmutation fuel cycles in TRU fueled reactors (without compromising safety), which may be essential for significant reduction in high-level waste repository requirements. ITER will serve as the prototype for the fusion neutron source, which means SABRs could be implemented to help close the nuclear fuel cycle during the 2^nd quarter of the century.

  15. Capital Cost: Pressurized Water Reactor Plant Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate.

  16. Waste management of tar water from pyrolysis and gasification of biomass in biogas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mogensen, A.S.; Schmidt, J.E.; Angelidaki, R.; Ahring, B.K.

    1998-08-01

    The digestion and detoxification of pyrolysis condensate and wet oxidised pyrolysis condensate was studied in different reactor systems: combined anaerobic and denitrifying UASB reactors, conventional UASB reactors and CSTR`s. The pyrolysis condensate and the wet oxidised condensate have a biogas potential of 190 m{sup 3}/ton VS, and the low amount of suspended solids is allowing the waste water to be treated in the UASB reactor as well as in the CSTR. The pyrolysis condensate could successfully be degraded in a CSTR in a 5% concentration when co-digested with manure, and the wet oxidised pyrolysis condensate could be degraded when added at a concentration of 30%. The UASB reactor was preferred over the CSTR since the xenobiotic compounds present in the waste water might easily be absorbed in the co-substrate required when using the CSTR technology. Consequently, decreased degradation of xenobiotics would be observed in the CSTR. A combined anaerobic and denitrifying UASB reactor was successfully digesting 5.5% of wet oxidised pyrolysis condensate, but further loading increments deteriorated the anaerobic digestion process. However, when a UASB reactor was fed with pyrolysis condensate (up to 100%) good reactor operation was observed indicating that the waste could be used as substrate in the biogas process, even in very high concentrations. The detoxification of pyrolysis condensate was further studied and the toxicity of pyrolysis condensate was decreased more than 77 times in the UASB reactor that was operating on 100% pyrolysis condensate. Phenol, methyl and dimethyl phenols along with methoxyphenols were shown to be degraded within the rector systems. Degradation rates for phenol and substituted phenols were determined indicating that the biomass was selective towards the substrates. Maximum growth rates and half saturation constants for phenol, 4-Methylphenol and 2-Methoxy-4-methylphenol were determined in batch experiments. A UASB reactor concept was further

  17. Students' Conceptions of Water Transport

    Science.gov (United States)

    Rundgren, Carl-Johan; Rundgren, Shu-Nu Chang; Schonborn, Konrad J.

    2010-01-01

    Understanding diffusion of water into and out of the cell through osmosis is fundamental to the learning and teaching of biology. Although this process is thought of as occurring directly across the lipid bilayer, the majority of water transport is actually mediated by specialised transmembrane water-channels called aquaporins. This study…

  18. Tar water digestion in an upflow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

    Skibsted Mogensen, A.; Angelidaki, I.; Schmidt, J.E.; Ahring, B.K. [Technical Univ., Dept. of Environmental Science and Engineering, Lyngby (Denmark)

    1998-08-01

    The water from the gasification and wet oxidised tar water has been digested anaerobically in UASB reactors and were digested in respectively 10 and 50% in batches. Though the tar water show inhibition at very low concentrations to aerobic microorganisms, the granular sludge used in UASB reactors degrades tar water in concentrations that reveal total inhibition of e.g. bacteria conducting the nitrification process. The value of waste waters are determined, showing that the tar water produces more biogas in the anaerobic digestion. A wide range of xenobiotics, especially phenolic compounds can be transformed in the anaerobic digestion process. Seven phenolic are followed in batch experiments and UASB reactor experiments, and their particular fate in the anaerobic systems embody large differences in the transformation pattern. (au) 24 refs.

  19. Multi-Application Small Light Water Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept

  20. Multi-Application Small Light Water Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Modro, S.M.; Fisher, J.E.; Weaver, K.D.; Reyes, J.N.; Groome, J.T.; Babka, P.; Carlson, T.M.

    2003-12-01

    The Multi-Application Small Light Water Reactor (MASLWR) project was conducted under the auspices of the Nuclear Energy Research Initiative (NERI) of the U.S. Department of Energy (DOE). The primary project objectives were to develop the conceptual design for a safe and economic small, natural circulation light water reactor, to address the economic and safety attributes of the concept, and to demonstrate the technical feasibility by testing in an integral test facility. This report presents the results of the project. After an initial exploratory and evolutionary process, as documented in the October 2000 report, the project focused on developing a modular reactor design that consists of a self-contained assembly with a reactor vessel, steam generators, and containment. These modular units would be manufactured at a single centralized facility, transported by rail, road, and/or ship, and installed as a series of self-contained units. This approach also allows for staged construction of an NPP and ''pull and replace'' refueling and maintenance during each five-year refueling cycle. Development of the baseline design concept has been sufficiently completed to determine that it complies with the safety requirements and criteria, and satisfies the major goals already noted. The more significant features of the baseline single-unit design concept include: (1) Thermal Power--150 MWt; (2) Net Electrical Output--35 MWe; (3) Steam Generator Type--Vertical, helical tubes; (4) Fuel UO{sub 2}, 8% enriched; (5) Refueling Intervals--5 years; (6) Life-Cycle--60 years. The economic performance was assessed by designing a power plant with an electric generation capacity in the range of current and advanced evolutionary systems. This approach allows for direct comparison of economic performance and forms a basis for further evaluation, economic and technical, of the proposed design and for the design evolution towards a more cost competitive concept

  1. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

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

  2. Antineutrino monitoring for the Iranian heavy water reactor

    CERN Document Server

    Christensen, Eric; Jaffke, Patrick; Shea, Thomas

    2014-01-01

    In this note we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a non-proliferation measure. We demonstrate that an above ground detector positioned right outside the IR-40 reactor building could meet and in some cases significantly exceed the verification goals identified by IAEA for plutonium production or diversion from declared inventories. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shutdown. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to re-establish continuity of knowledge in safeguards conclusions should this become necessary.

  3. Light Water Reactor Sustainability Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  4. Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

    Energy Technology Data Exchange (ETDEWEB)

    Ammirabile, Luca, E-mail: luca.ammirabile@ec.europa.e [European Commission, JRC, Institute for Energy, Westerduinweg 3, 1755 LE Petten (Netherlands)

    2010-10-15

    In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).

  5. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  6. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  7. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  8. Corrosion and Corrosion Control in Light Water Reactors

    Science.gov (United States)

    Gordon, Barry M.

    2013-08-01

    Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

  9. Core management of the prototype heavy water reactor FUGEN

    Energy Technology Data Exchange (ETDEWEB)

    Deshimaru, Takehide; Furubayashi, Toshiyuki; Matsumoto, Mitsuo (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

    1982-12-01

    In this paper, the core management which has been implemented so far for the prototype heavy water reactor FUGEN is described. First, the outline of the core is introduced. The core management is generally the repetition of planning, practice and evaluation, but the evaluation is specifically important in FUGEN because FUGEN is a prototype reactor. In the reactor FUGEN, the fuel replacement plan which determines the number and position of fuels to be replaced, and fuel procurement plan based on the replacement plan are prepared. The control rod pattern is determined so that the thermal limit for the fuel assembly is secured throughout the fuel cycle, but the output flattening by control rods is scarcely necessary by adopting a distributed replacement method. After a replaced core has been composed, the maximum excess reactivity and reactivity shut-down margin are mainly measured at the start-up of the reactor to confirm the predetermined characteristics of the replaced core. The core life can be simply and accurately estimated by the measurement of /sup 10/B concentration in heavy water. The output distribution in the core is an important parameter for calculating the performance of the FUGEN reactor core. The output increasing procedure is also controlled in accordance with that of light water reactors.

  10. Materials technology for an advanced space power nuclear reactor concept: Program summary

    Science.gov (United States)

    Gluyas, R. E.; Watson, G. K.

    1975-01-01

    The results of a materials technology program for a long-life (50,000 hr), high-temperature (950 C coolant outlet), lithium-cooled, nuclear space power reactor concept are reviewed and discussed. Fabrication methods and compatibility and property data were developed for candidate materials for fuel pins and, to a lesser extent, for potential control systems, reflectors, reactor vessel and piping, and other reactor structural materials. The effects of selected materials variables on fuel pin irradiation performance were determined. The most promising materials for fuel pins were found to be 85 percent dense uranium mononitride (UN) fuel clad with tungsten-lined T-111 (Ta-8W-2Hf).

  11. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

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

    Science.gov (United States)

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

    2016-02-01

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; 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)

    2016-02-08

    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.

  14. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  15. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  16. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    OpenAIRE

    Masaki Kurata

    2016-01-01

    Research and development (R&D) methodology for the practical use of accident tolerant fuel (ATF) in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening proces...

  17. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    OpenAIRE

    Hyun-Gil Kim; Jae-Ho Yang; Weon-Ju Kim; Yang-Hyun Koo

    2016-01-01

    For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density compo...

  18. Chemical Gradients in Crud on Boiling Water Reactor Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    D. L. Porter; D. E. Janney

    2007-04-01

    Crud (radioactive corrosion products formed inside nuclear reactors is a major problem in commercial power-producing nuclear reactors. Although there are numerous studies of simulated (non-radioactive) crud, characteristics of crud from actual reactors are rarely studied. This study reports scanning electron microscope (SEM) studies of fragments of crud from a commercially operating boiling water reactor. Chemical analyses in the SEM indicated that the crud closest to the outer surfaces of the fuel pins in some areas had Fe:Zn ratios close to 2:1, which decreased away from the fuel pin in some of the fragments. In combination with transmission electron microsope analyses (published elsewhere), these results suggest that the innermost layer of crud in some areas may consist of franklinite (ZnFe2O4, also called zinc spinel), while outer layers in these areas may be predominantly iron oxides.

  19. High-Temperature Water-Gas Shift Membrane Reactor Study

    Energy Technology Data Exchange (ETDEWEB)

    Ciocco, M.V.; Iyoha, O.; Enick, R.M.; Killmeyer, R.P.

    2007-06-01

    NETL’s Office of Research and Development is exploring the integration of membrane reactors into coal gasification plants as a way of increasing efficiency and reducing costs. Water-Gas Shift Reaction experiments were conducted in membrane reactors at conditions similar to those encountered at the outlet of a coal gasifier. The changes in reactant conversion and product selectivity due to the removal of hydrogen via the membrane reactor were quantified. Research was conducted to determine the influence of residence time and H2S on CO conversion in both Pd and Pd80wt%Cu membrane reactors. Effects of the hydrogen sulfide-to-hydrogen ratio on palladium and a palladium-copper alloy at high-temperature were also investigated. These results were compared to thermodynamic calculations for the stability of palladium sulfides.

  20. Preliminary Evaluation of Convective Heat Transfer in a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson J. Boise; Reid, Robert S.

    2007-01-01

    As part of the Vision for Space Exploration, the end of the next decade will bring man back to the surface of the moon. A crucial issue for the establishment of human presence on the moon will be the availability of compact power sources. This presence could require greater than 10's of kWt's in follow on years. Nuclear reactors are well suited to meet the needs for power generation on the lunar or Martian surface. Radiation shielding is a key component of any surface power reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), and boron carbide. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to fix the location of any vapor that could form radiation streaming paths. The water shield concept relies on the predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. NASA Marshall Space Flight Center has developed the experience and facilities necessary to do this evaluation in its Early Flight Fission - Test Facility (EFF-TF).

  1. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  2. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    Energy Technology Data Exchange (ETDEWEB)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  3. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  4. Evaluation of fatigue data including reactor water environmental effects

    Energy Technology Data Exchange (ETDEWEB)

    Rosinski, S.T. [EPRI, Charlotte, NC (United States); Nickell, R.E. [Applied Science and Technology, Poway, CA (United States); Van Der Sluys, W.A. [Alliance, OH (United States); Yukawa, S. [Boulder, CO (United States)

    2002-07-01

    Laboratory data have been gathered in the past decade indicating a significant reduction in component fatigue life when reactor water environmental effects are experimentally simulated. However, these laboratory data have not been supported by nuclear power plant component operating experience. The laboratory data under simulated operating conditions are being used to support arguments for revising the design-basis fatigue curves in the ASME Code Section III, Division 1, for Class 1 components. A thorough review of available laboratory fatigue data and their applicability to actual component operating conditions was performed. The evaluation divided the assembly, review and assessment of existing laboratory fatigue data and its applicability to plant operating conditions into four principal tasks: (1) review of available laboratory data relative to thresholds for environmental parameters, such as temperature, reactor water oxidation potential, strain rate, strain amplitude, reactor water flow rate, and component metal sulfur content; (2) determination of the relevance of the laboratory data to actual plant operating conditions; (3) review of laboratory S-N data curve-fitting models; and (4) assessment of existing ASME Code Section III Class 1 margins This paper summarizes the results of the data review. In addition, recommendations are made for additional laboratory testing intended to improve the applicability of laboratory test results under simulated reactor water environmental conditions. (authors)

  5. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

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

    2015-04-01

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

  6. Experimental investigation of a directionally enhanced DHX concept for high temperature Direct Reactor Auxiliary Cooling Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, Joel T.; Blandford, Edward D., E-mail: edb@unm.edu

    2016-07-15

    Highlights: • A novel directional heat exchanger design has been developed. • Hydrodynamic tests have been performed on the proposed design. • Heat transfer performance is inferred by hydrodynamic results. • Results are discussed and future work is suggested. - Abstract: The use of Direct Reactor Auxiliary Cooling Systems (DRACSs) as a safety-related decay heat removal system for advanced reactors has developed historically through the Sodium Fast Reactor (SFR) community. Beginning with the EBR-II, DRACSs have been utilized in a large number of past and current SFR designs. More recently, the DRACS has been adopted for Fluoride Salt-Cooled High-Temperature Reactors (FHRs) for similar decay heat removal functions. In this paper we introduce a novel directionally enhanced DRACS Heat Exchanger (DHX) concept. We present design options for optimizing such a heat exchanger so that shell-side heat transfer is enhanced in one primary coolant flow direction and degraded in the opposite coolant flow direction. A reduced-scale experiment investigating the hydrodynamics of a directionally enhanced DHX was built and the data collected is presented. The concept of thermal diodicity is expanded to heat exchanger technologies and used as performance criteria for evaluating design options. A heat exchanger that can perform as such would be advantageous for use in advanced reactor concepts where primary coolant flow reversal is expected during Loss-of-Forced-Circulation (LOFC) accidents where the ability to circulate coolant is compromised. The design could also find potential use in certain advanced Sodium Fast Reactor (SFR) designs utilizing fluidic diode concepts.

  7. Integral Inherently Safe Light Water Reactor (I2S-LWR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, Bojan [Georgia Inst. of Technology, Atlanta, GA (United States); Memmott, Matthew [Brigham Young Univ., Provo, UT (United States); Boy, Guy [Florida Inst. of Technology, Melbourne, FL (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Lee, John [Univ. of Michigan, Ann Arbor, MI (United States); Muldrow, Lycurgus [Morehouse College, Atlanta, GA (United States); Upadhyaya, Belle [Univ. of Tennessee, Knoxville, TN (United States); Hines, Wesley [Univ. of Tennessee, Knoxville, TN (United States); Haghighat, Alierza [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States)

    2017-10-02

    This final report summarizes results of the multi-year effort performed during the period 2/2013- 12/2016 under the DOE NEUP IRP Project “Integral Inherently Safe Light Water Reactors (I2S-LWR)”. The goal of the project was to develop a concept of a 1 GWe PWR with integral configuration and inherent safety features, at the same time accounting for lessons learned from the Fukushima accident, and keeping in mind the economic viability of the new concept. Essentially (see Figure 1-1) the project aimed to implement attractive safety features, typically found only in SMRs, to a larger power (1 GWe) reactor, to address the preference of some utilities in the US power market for unit power level on the order of 1 GWe.

  8. Materials Degradation in Light Water Reactors: Life After 60,???

    Energy Technology Data Exchange (ETDEWEB)

    Busby, Jeremy T [ORNL; Nanstad, Randy K [ORNL; Stoller, Roger E [ORNL; Feng, Zhili [ORNL; Naus, Dan J [ORNL

    2008-04-01

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase

  9. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  10. Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klein, Andrew [Oregon State Univ., Corvallis, OR (United States). Nuclear Engineering and Radiation Health Physics; Matthews, Topher [Oregon State Univ., Corvallis, OR (United States); Lenhof, Renae [Oregon State Univ., Corvallis, OR (United States); Deason, Wesley [Oregon State Univ., Corvallis, OR (United States); Harter, Jackson [Oregon State Univ., Corvallis, OR (United States)

    2015-01-16

    Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

  11. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.''...

  12. 78 FR 56752 - Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors...

    Science.gov (United States)

    2013-09-13

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Interim Staff Guidance Specific Environmental Guidance for Integral Pressurized Water Reactors... and operate integral pressurized water reactors (iPWR). This guidance applies to environmental reviews...

  13. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR) standard...

  14. Use of reactor effluent water as steam plant boiler feed

    Energy Technology Data Exchange (ETDEWEB)

    Clukey, H.V.

    1953-12-08

    The radiological aspects of a proposal to recover some of the heat now wasted in cooling water from the Hanford reactors by using the hot water as boiler feed for the steam plants in the 100 Areas are evaluated. The radioactive material in the hot effluent water will contaminate the boiler feed water system, cause additional radiation exposure of personnel, and increase the cost of maintenance and radiation protection, but very little radioactive material will be carried over into the steam system. At present steam loads, this proposal is economically attractive; other proposals being considered may nullify any savings from this one. 21 refs., 1 fig., 10 tabs.

  15. A new reactor concept for sludge reduction using aquatic worms

    NARCIS (Netherlands)

    Elissen, H.J.H.; Hendrickx, T.L.G.; Temmink, B.G.; Buisman, C.J.N.

    2006-01-01

    Biological waste water treatment results in the production of waste sludge. The final treatment option in The Netherlands for this waste sludge is usually incineration. A biological approach to reduce the amount of waste sludge is through predation by aquatic worms. In this paper we test the applica

  16. A new reactor concept for sludge reduction using aquatic worms

    NARCIS (Netherlands)

    Elissen, H.J.H.; Hendrickx, T.L.G.; Temmink, B.G.; Buisman, C.J.N.

    2006-01-01

    Biological waste water treatment results in the production of waste sludge. The final treatment option in The Netherlands for this waste sludge is usually incineration. A biological approach to reduce the amount of waste sludge is through predation by aquatic worms. In this paper we test the

  17. Reactor

    Science.gov (United States)

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  18. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  19. Upper internals arrangement for a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Singleton, Norman R; Altman, David A; Yu, Ching; Rex, James A; Forsyth, David R

    2013-07-09

    In a pressurized water reactor with all of the in-core instrumentation gaining access to the core through the reactor head, each fuel assembly in which the instrumentation is introduced is aligned with an upper internals instrumentation guide-way. In the elevations above the upper internals upper support assembly, the instrumentation is protected and aligned by upper mounted instrumentation columns that are part of the instrumentation guide-way and extend from the upper support assembly towards the reactor head in hue with a corresponding head penetration. The upper mounted instrumentation columns are supported laterally at one end by an upper guide tube and at the other end by the upper support plate.

  20. COMSORS: A light water reactor chemical core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States). Chemical Technology Div.; Kenton, M.A. [Creare Inc., Hanover, NH (United States)

    1997-02-24

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B{sub 2}O{sub 3}) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation.

  1. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

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

  2. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

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

  3. 77 FR 38338 - Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative; La Crosse Boiling Water Reactor Exemption From Certain Security Requirements 1.0 Background The La Crosse Boiling Water Reactor (LACBWR) is owned and was operated by the..., which utilized a forced-circulation, direct-cycle boiling water reactor as its heat source. The plant is...

  4. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

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

  5. Concept of an inherently-safe high temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

    2012-06-06

    As the challenge to ensure no harmful release of radioactive materials at the accidents by deterministic approach instead to satisfy acceptance criteria or safety goal for risk by probabilistic approach, new concept of advanced reactor, an inherently-safe high temperature gas-cooled reactor, is proposed based on the experience of the operation of the actual High Temperature Gas-cooled Reactor (HTGR) in Japan, High Temperature Engineering Test Reactor (HTTR), and the design of the commercial plant (GTHTR300), utilizing the inherent safety features of the HTGR (i.e., safety features based on physical phenomena). The safety design philosophy of the inherently-safe HTGR for the safety analysis of the radiological consequences is determined as the confinement of radioactive materials is assured by only inherent safety features without engineered safety features, AC power or prompt actions by plant personnel if the design extension conditions occur. Inherent safety features to prevent the loss or degradation of the confinement function are identified. It is proposed not to apply the probabilistic approach for the evaluation of the radiological consequences of the accidents in the safety analysis because no inherent safety features fail for the mitigation of the consequences of the accidents. Consequently, there are no event sequences to harmful release of radioactive materials if the design extension conditions occur in the inherently-safe HTGR concept. The concept and future R and D items for the inherently-safe HTGR are described in this paper.

  6. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  7. Optofluidic planar reactors for photocatalytic water treatment using solar energy

    Science.gov (United States)

    Lei, Lei; Wang, Ning; Zhang, X. M.; Tai, Qidong; Tsai, Din Ping; Chan, Helen L. W.

    2010-01-01

    Optofluidics may hold the key to greater success of photocatalytic water treatment. This is evidenced by our findings in this paper that the planar microfluidic reactor can overcome the limitations of mass transfer and photon transfer in the previous photocatalytic reactors and improve the photoreaction efficiency by more than 100 times. The microreactor has a planar chamber (5 cm×1.8 cm×100 μm) enclosed by two TiO2-coated glass slides as the top cover and bottom substrate and a microstructured UV-cured NOA81 layer as the sealant and flow input∕output. In experiment, the microreactor achieves 30% degradation of 3 ml 3×10−5M methylene blue within 5 min and shows a reaction rate constant two orders higher than the bulk reactor. Under optimized conditions, a reaction rate of 8% s−1 is achieved under solar irradiation. The average apparent quantum efficiency is found to be only 0.25%, but the effective apparent quantum efficiency reaches as high as 25%. Optofluidic reactors inherit the merits of microfluidics, such as large surface∕volume ratio, easy flow control, and rapid fabrication and offer a promising prospect for large-volume photocatalytic water treatment. PMID:21267436

  8. Gravity Scaling of a Power Reactor Water Shield

    Science.gov (United States)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for use on initial lunar surface power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxiliary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2007). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa(sup n). These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined.

  9. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

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

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    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.

  11. An autonomous long-term fast reactor system and the principal design limitations of the concept

    Science.gov (United States)

    Tsvetkova, Galina Valeryevna

    The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National

  12. Spiral-shaped reactor for water disinfection

    KAUST Repository

    Soukane, Sofiane

    2016-04-20

    Chlorine-based processes are still widely used for water disinfection. The disinfection process for municipal water consumption is usually carried out in large tanks, specifically designed to verify several hydraulic and disinfection criteria. The hydrodynamic behavior of contact tanks of different shapes, each with an approximate total volume of 50,000 m3, was analyzed by solving turbulent momentum transport equations with a computational fluid dynamics code, namely ANSYS fluent. Numerical experiments of a tracer pulse were performed for each design to generate flow through curves and investigate species residence time distribution for different inlet flow rates, ranging from 3 to 12 m3 s−1. A new nature-inspired Conch tank design whose shape follows an Archimedean spiral was then developed. The spiral design is shown to strongly outperform the other tanks’ designs for all the selected plug flow criteria with an enhancement in efficiency, less short circuiting, and an order of magnitude improvement in mixing and dispersion. Moreover, following the intensification philosophy, after 50% reduction in its size, the new design retains its properties and still gives far better results than the classical shapes.

  13. Dense Medium Plasma Water Purification Reactor (DMP WaPR) Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The Dense Medium Plasma Water Purification Reactor offers significant improvements over existing water purification technologies used in Advanced Life Support...

  14. Water security assessment using blue and green water footprint concepts

    Science.gov (United States)

    Veettil, Anoop Valiya; Mishra, Ashok K.

    2016-11-01

    The quantitative assessment of water security using the concept of blue and green water footprints can improve water resources management at local to regional scale. We developed an integrated modeling framework by considering both climatic and anthropogenic factors to investigate spatio-temporal variability of blue and green water availability and to quantify the water security in a river basin. The proposed modeling framework can be useful for providing an overview of the water security within the watershed and to identify water stress (hot spots) regions within the river basin. We applied Soil and Water Assessment Tool (SWAT) to quantify the availability of fresh water (blue water and green water) in Savannah River Basin (SRB), USA. The anthropogenic factors (e.g., water demand) and Environmental Flow Requirement (EFR) information are incorporated to quantify the water security in terms of scarcity and vulnerability indices. A higher amount of blue water was observed for counties located in the upper part of SRB and higher green water flow was observed for counties that has the presence of intensive agriculture and large water bodies (e.g., reservoir). A time lag exists between the maximum rainfall during June-September and the maximum blue water observed in December-March. The study also analyzed the monthly variation of blue and green water flow for counties located in SRB. We expect that the water security assessment can provide useful information for understanding the emerging hot spots within a river basin (eco-system) due to the abstraction of water for human activities, such as irrigation, industrial use, energy production and domestic use.

  15. Wear resistant zirconium base alloy article for water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gillett, J.E.; Shockling, L.A.; Sherwood, D.G.

    1988-03-01

    In a water reactor operating environment, the combination having improved fretting wear resistance is described comprising: an elongated tubular water displacer rod; having a low neutron absorption cross section guide support plates distributed along the length of the water displacer rod; the water displacer rod intersecting the guide support plates through apertures in the guide support plates; the water displacer rod having a plurality of spaced apart annular electrospark deposited coatings, each coating facing the wall of a respective aperture, the electrospark deposited coatings comprising Cr/sub 2/C/sub 3/; wherein the water displacer rod has a tube wall composed of a zirconium base alloy; and wherein the guide support plates are composed of a stainless steel alloy.

  16. A HUMAN AUTOMATION INTERACTION CONCEPT FOR A SMALL MODULAR REACTOR CONTROL ROOM

    Energy Technology Data Exchange (ETDEWEB)

    Le Blanc, Katya; Spielman, Zach; Hill, Rachael

    2017-06-01

    Many advanced nuclear power plant (NPP) designs incorporate higher degrees of automation than the existing fleet of NPPs. Automation is being introduced or proposed in NPPs through a wide variety of systems and technologies, such as advanced displays, computer-based procedures, advanced alarm systems, and computerized operator support systems. Additionally, many new reactor concepts, both full scale and small modular reactors, are proposing increased automation and reduced staffing as part of their concept of operations. However, research consistently finds that there is a fundamental tradeoff between system performance with increased automation and reduced human performance. There is a need to address the question of how to achieve high performance and efficiency of high levels of automation without degrading human performance. One example of a new NPP concept that will utilize greater degrees of automation is the SMR concept from NuScale Power. The NuScale Power design requires 12 modular units to be operated in one single control room, which leads to a need for higher degrees of automation in the control room. Idaho National Laboratory (INL) researchers and NuScale Power human factors and operations staff are working on a collaborative project to address the human performance challenges of increased automation and to determine the principles that lead to optimal performance in highly automated systems. This paper will describe this concept in detail and will describe an experimental test of the concept. The benefits and challenges of the approach will be discussed.

  17. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  18. Operational limitations of light water reactors relating to fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, H S

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed.

  19. Water-moderated reactor fuel cladding reliability study

    OpenAIRE

    Бакутяк, Елена Викторовна; Пелых, Сергей Николаевич

    2014-01-01

    Considering the fuel element, averaged by fuel assembly (FA) of water-moderated reactor with the power of 1000 MW (VVER-1000), the number of fuel elements with the greatest cladding failure probability after 4 operation years at Khmelnitsky NPP-2 (KNPP-2) is found. This will allow to calculate the fuel cladding failure probability and determine the most likely cladding damages, which will enable to improve the performance and economic indexes of VVER.The novelty of the paper lies in calculati...

  20. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  1. LIGHT WATER REACTOR ACCIDENT TOLERANT FUELS IRRADIATION TESTING

    Energy Technology Data Exchange (ETDEWEB)

    Carmack, William Jonathan [Idaho National Laboratory; Barrett, Kristine Eloise [Idaho National Laboratory; Chichester, Heather Jean MacLean [Idaho National Laboratory

    2015-09-01

    The purpose of Accident Tolerant Fuels (ATF) experiments is to test novel fuel and cladding concepts designed to replace the current zirconium alloy uranium dioxide (UO2) fuel system. The objective of this Research and Development (R&D) is to develop novel ATF concepts that will be able to withstand loss of active cooling in the reactor core for a considerably longer time period than the current fuel system while maintaining or improving the fuel performance during normal operations, operational transients, design basis, and beyond design basis events. It was necessary to design, analyze, and fabricate drop-in capsules to meet the requirements for testing under prototypic LWR temperatures in Idaho National Laboratory's Advanced Test Reactor (ATR). Three industry led teams and one DOE team from Oak Ridge National Laboratory provided fuel rodlet samples for their new concepts for ATR insertion in 2015. As-built projected temperature calculations were performed on the ATF capsules using the BISON fuel performance code. BISON is an application of INL’s Multi-physics Object Oriented Simulation Environment (MOOSE), which is a massively parallel finite element based framework used to solve systems of fully coupled nonlinear partial differential equations. Both 2D and 3D models were set up to examine cladding and fuel performance.

  2. A breed and burn reshuffling scheme for an Astrid-like reactor concept design

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois L, J. L., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2016-09-15

    The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

  3. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems

    NARCIS (Netherlands)

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.

    In

  4. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems.

    NARCIS (Netherlands)

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.In this the

  5. Automated disposable small scale reactor for high throughput bioprocess development: a proof of concept study.

    Science.gov (United States)

    Bareither, Rachel; Bargh, Neil; Oakeshott, Robert; Watts, Kathryn; Pollard, David

    2013-12-01

    The acceleration of bioprocess development for biologics and vaccines can be enabled by automated high throughput technologies. This will alleviate the significant resource burden from the multi-factorial statistical experimentation required for controlling product quality attributes of complex biologics. Recent technology advances have improved clone evaluation and screening, but have struggled to combine the scale down criteria required for both high cell density cell culture and microbial processes, with sufficient automation and disposable technologies to accelerate process development. This article describes the proof of concept evaluations of an automated disposable small scale reactor for high throughput upstream process development. Characterization studies established the small scale stirred tank disposable 250 mL reactor as similar to those of lab and pilot scale. The reactor generated equivalent process performance for industrial biologics processes for therapeutic protein and monoclonal antibody production using CHO cell culture, Pichia pastoris and E. coli. This included similar growth, cell viability, product titer, and product quality. The technology was shown to be robust across multiple runs and met the requirements for the ability to run high cell density processes (>400 g/L wet cell weight) with exponential feeds and sophisticated event triggered processes. Combining this reactor into an automated array of reactors will ultimately be part of a high throughput process development strategy. This will combine upstream, small scale purification with rapid analytics that will dramatically shorten timelines and costs of developing biological processes.

  6. Reactor concepts for atomic layer deposition on agitated particles: A review

    Energy Technology Data Exchange (ETDEWEB)

    Longrie, Delphine, E-mail: delphine.longrie@asm.com; Deduytsche, Davy; Detavernier, Christophe, E-mail: christophe.detavernier@ugent.be [Department of Solid State Sciences, Ghent University, Krijgslaan 281/S1, B-9000 Gent (Belgium)

    2014-01-15

    The number of possible applications for nanoparticles has strongly increased in the last decade. For many applications, nanoparticles with different surface and bulk properties are necessary. A popular surface modification technique is coating the particle surface with a nanometer thick layer. Atomic layer deposition (ALD) is known as a reliable method for depositing ultrathin and conformal coatings. In this article, agitation or fluidization of the particles is necessary for performing ALD on (nano)particles. The principles of gas fluidization of particles will be outlined, and a classification of the gas fluidization behavior of particles based on their size and density will be given. Following different reactor concepts that have been designed to conformally coat (nano)particles with ALD will be described, and a concise overview will be presented of the work that has been performed with each of them ending with a concept reactor for performing spatial ALD on fluidized particles.

  7. Numerical study of the effects of lamp configuration and reactor wall roughness in an open channel water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu

    2016-07-01

    This article describes the assessment of a numerical procedure used to determine the UV lamp configuration and surface roughness effects on an open channel water disinfection UV reactor. The performance of the open channel water disinfection UV reactor was numerically analyzed on the basis of the performance indictor reduction equivalent dose (RED). The RED values were calculated as a function of the Reynolds number to monitor the performance. The flow through the open channel UV reactor was modelled using a k-ε model with scalable wall function, a discrete ordinate (DO) model for fluence rate calculation, a volume of fluid (VOF) model to locate the unknown free surface, a discrete phase model (DPM) to track the pathogen transport, and a modified law of the wall to incorporate the reactor wall roughness effects. The performance analysis was carried out using commercial CFD software (ANSYS Fluent 15.0). Four case studies were analyzed based on open channel UV reactor type (horizontal and vertical) and lamp configuration (parallel and staggered). The results show that lamp configuration can play an important role in the performance of an open channel water disinfection UV reactor. The effects of the reactor wall roughness were Reynolds number dependent. The proposed methodology is useful for performance optimization of an open channel water disinfection UV reactor.

  8. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR...

  9. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems

    OpenAIRE

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.

    In this thesis the possibilities are presented for fixed-film post-treatment of anaerobically digested domestic sewage and water reconditioning in aquacultural water recirculation systems. Emphasis i...

  10. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  11. Simulation of Water Gas Shift Zeolite Membrane Reactor

    Science.gov (United States)

    Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

    2017-07-01

    The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

  12. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  13. Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

    Energy Technology Data Exchange (ETDEWEB)

    Pope, Michael A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ryskamp, John M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2013-09-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a

  14. Non-linear analysis in Light Water Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  15. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

    Cox, B. [University of Toronto, Centre for Nuclear Engineering, Toronto, Ontario (Canada)

    1997-07-01

    Water chemistry may have a major impact on fuel cladding performance in PWRs. If the saturation temperature on the surface of fuel cladding is exceeded, either because of the thermal hydraulics of the system, or because of crud deposition, then LiOH concentration can occur within thick porous oxide films on the cladding. This can degrade the protective film and accelerate the corrosion rate of the cladding. If sufficient boric acid is also present in the coolant then these effects may be mitigated. This is normally the case through most of any reactor fuel cycle. Extensive surface boiling may disrupt this equilibrium because of the volatility of boric acid in steam. Under such conditions severe cladding corrosion can ensue. The potential for such effects on high burnup cladding in CANDU reactors, where bone acid is not present in the primary coolant, is discussed. (author)

  16. Transactions of the nineteenth water reactor safety information meeting

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. (comp.)

    1991-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 19th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 28--30, 1991. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from the governments and industry in Europe and Japan are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting, and are given in the order of their presentation in each session. The individual summaries have been cataloged separately.

  17. Proceedings of the Water Reactor Fuel Performance Meeting - WRFPM / Top Fuel 2009

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2009-06-15

    SFEN, ENS, SNR, ANS, AESJ, CNS KNS, IAEA and NEA are jointly organizing the 2009 International Water Reactor Fuel Performance / TopFuel 2009 Meeting following the 2008 KNS Water Reactor Performance Meeting held during October 19-23, 2008 in Seoul, Korea. This meeting is held annually on a tri-annual rotational basis in Europe, USA and Asia. In 2009, this meeting will be held in Paris, September 6-10, 2009 in coordination with the Global 2009 Conference at the same date and place. That would lead to a common opening session, some common technical presentations, a common exhibition and common social events. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as manufacturing, performance in commercial and test reactors or on-going and future developments and trends. Emphasis will be placed on fuel reliability in the general context of nuclear 'Renaissance' and recycling perspective. The meeting includes selectively front and/or back end issues that impact fuel designs and performance. In this frame, the conference track devoted to 'Concepts for transportation and interim storage of spent fuels and conditioned waste' will be shared with 'GLOBAL' conference. Technical Tracks: - 1. Fuel Performance, Reliability and Operational Experience: Fuel operating experience and performance; experience with high burn-up fuels; water side corrosion; stress corrosion cracking; MOX fuel performance; post irradiation data on lead fuel assemblies; radiation effects; water chemistry and corrosion counter-measures. - 2. Transient Fuel Behaviour and Safety Related Issues: Transient fuel behavior and criteria (RIA, LOCA, ATWS, Ramp tests..). Fuel safety-related issues such as PCI (pellet cladding interaction), transient fission gas releases and cladding bursting/ballooning during transient events - Advances in fuel performance modeling and core reload methodology, small and large-scale fuel testing

  18. Materials Inventory Database for the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Kazi Ahmed; Shannon M. Bragg-Sitton

    2013-08-01

    Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

  19. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  20. Numerical study of the effects of surface roughness on water disinfection UV reactor.

    Science.gov (United States)

    Sultan, Tipu; Ahmad, Sarfraz; Cho, Jinsoo

    2016-04-01

    UV reactors are an emerging choice as a big barrier against the pathogens present in drinking water. However, the precise role of reactor's wall roughness for cross flow ultraviolet (CF-UV) and axial flow ultraviolet (AF-UV) water disinfection reactors are unknown. In this paper, the influences of reactor's wall roughness were investigated with a view to identify their role on the performance factors namely dose distribution and reduction equivalent dose (RED). Herein, the relative effects of reactor's wall roughness on the performance of CF-UV and AF-UV reactors were also highlighted. This numerical study is a first step towards the comprehensive analysis of the effects of reactor's wall roughness for UV reactor. A numerical analysis was performed using ANSYS Fluent 15 academic version. The reactor's wall roughness has a significant effect on the RED. We found that the increase in RED is Reynolds number dependent (at lower value of turbulent Reynolds number the effects are remarkable). The effects of reactor's roughness were more pronounced for AF-UV reactor. The simulation results suggest that the study of reactor's wall roughness provides valuable insight to fully understand the effects of reactor's wall roughness and its impact on the flow behavior and other features of CF-UV and AF-UV water disinfection reactors. Copyright © 2016 Elsevier Ltd. All rights reserved.

  1. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  2. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Science.gov (United States)

    Anuar, Nuraslinda; Muhamad Pauzi, Anas

    2016-01-01

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the βmin is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the βmin, resulting in a list of candidate designs that possess the β value that is larger than the βmin. The proposed methodology can also be applied to purposes other than technological foresight.

  3. Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

    Science.gov (United States)

    Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

    2017-01-01

    The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

  4. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Energy Technology Data Exchange (ETDEWEB)

    Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.

  5. In-reactor tests of the nuclear light bulb rocket concept

    Science.gov (United States)

    Gauntt, R. O.; Slutz, S. A.; Latham, T. S.; Roman, W. C.; Rogers, R. J.

    1992-07-01

    An overview is given of the closed-cycle Gas Core Nuclear Rocket outlining scenarios for its use in short-duration Mars missions and results of Nuclear Light Bulb (NLB) tests. Isothermal and nonnuclear tests are described which confirmed the fundamental concepts behind the NLB. NLB reference-engine performance characteristics are given for hypothetical engines that could be used for manned Mars missions. Vehicle/propulsion sizing is based on a Mars mission with three trans-Mars impulse burns, capture and escape burns, and a total mission duration of 600 days. The engine would have a specific impulse of 1870 seconds, a 412-kN thrust, and a thrust/weight ratio of 1.3. Reactor tests including small-scale in-reactor tests are shown to be prerequisites for studying: (1) fluid mechanical confinement of the gaseous nuclear fuel; (2) buffer gas separation and circulation; and (3) the minimization of transparent wall-heat loading. The reactor tests are shown to be critical for establishing the feasibility of the NLB concept.

  6. Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-04-01

    The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

  7. Camera Inspection Arm for Boiling Water Reactors - 13330

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Scott; Rood, Marc [S.A. Technology, 3985 S. Lincoln Ave, Loveland, CO 80537 (United States)

    2013-07-01

    Boiling Water Reactor (BWR) outage maintenance tasks can be time-consuming and hazardous. Reactor facilities are continuously looking for quicker, safer, and more effective methods of performing routine inspection during these outages. In 2011, S.A. Technology (SAT) was approached by Energy Northwest to provide a remote system capable of increasing efficiencies related to Reactor Pressure Vessel (RPV) internal inspection activities. The specific intent of the system discussed was to inspect recirculation jet pumps in a manner that did not require manual tooling, and could be performed independently of other ongoing inspection activities. In 2012, SAT developed a compact, remote, camera inspection arm to create a safer, more efficient outage environment. This arm incorporates a compact and lightweight design along with the innovative use of bi-stable composite tubes to provide a six-degree of freedom inspection tool capable of reducing dose uptake, reducing crew size, and reducing the overall critical path for jet pump inspections. The prototype camera inspection arm unit is scheduled for final testing in early 2013 in preparation for the Columbia Generating Station refueling outage in the spring of 2013. (authors)

  8. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  9. Supercritical Water Reactor Cycle for Medium Power Applications

    Energy Technology Data Exchange (ETDEWEB)

    BD Middleton; J Buongiorno

    2007-04-25

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency {ge}20%; Steam turbine outlet quality {ge}90%; and Pumping power {le}2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  10. Characterization of 14C in Swedish light water reactors.

    Science.gov (United States)

    Magnusson, Asa; Aronsson, Per-Olof; Lundgren, Klas; Stenström, Kristina

    2008-08-01

    This paper presents the results of a 4-y investigation of 14C in different waste streams of both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Due to the potential impact of 14C on human health, minimizing waste and releases from the nuclear power industry is of considerable interest. The experimental data and conclusions may be implemented to select appropriate waste management strategies and practices at reactor units and disposal facilities. Organic and inorganic 14C in spent ion exchange resins, process water systems, ejector off-gas and replaced steam generator tubes were analyzed using a recently developed extraction method. Separate analysis of the chemical species is of importance in order to model and predict the fate of 14C within process systems as well as in dose calculations for disposal facilities. By combining the results of this investigation with newly calculated production rates, mass balance assessments were made of the 14C originating from production in the coolant. Of the 14C formed in the coolant of BWRs, 0.6-0.8% was found to be accumulated in the ion exchange resins (core-specific production rate in the coolant of a 2,500 MWth BWR calculated to be 580 GBq GW(e)(-1) y(-1)). The corresponding value for PWRs was 6-10% (production rate in a 2,775 MWth PWR calculated to be 350 GBq GW(e)(-1) y(-1)). The 14C released with liquid discharges was found to be insignificant, constituting less than 0.5% of the production in the coolant. The stack releases, routinely measured at the power plants, were found to correspond to 60-155% of the calculated coolant production, with large variations between the BWR units.

  11. Boiling-Water Reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-12-01

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

  13. Structural Integrity of Water Reactor Pressure Boundary Components.

    Science.gov (United States)

    1980-08-01

    tests of reference steels of the NRC light water reactor, pressure vessel irradiation dosimetry program. SECURITY CLAS5IICATION 0PHiS PA6GMbn" Dfat ...multiple specimen R- curve approach; NRL emphasis was on the SSC procedure as it is being developed for hot- cell testing of irradiated materials. MULTIPLE...a second autoclave, capable of testing 50 or 100 mm (2T or 4T) thick CT or WOL specimens, was installed in a hot cell and a test was started on 2T-CT

  14. Concepts and Applications of Water Transport Time Scales for Coastal Inlet Systems

    Science.gov (United States)

    2010-08-01

    the assumptions made. It measures the average time required to flush a system. Based on the concept of a continuous stirred tank reactor ( CSTR ), an...steady state. Flushing time calculated by Equations (1) or (2), or by the CSTR solution in Equation (4) tends to underestimate the time that a water...considering only tidal flushing without the return flow factor, the flushing time is 61.4 hr (2.6 days). They also applied the CSTR tracer method

  15. The hybrid reactor project based on the straight field line mirror concept

    Science.gov (United States)

    Ågren, O.; Noack, K.; Moiseenko, V. E.; Hagnestâl, A.; Källne, J.; Anglart, H.

    2012-06-01

    The straight field line mirror (SFLM) concept is aiming towards a steady-state compact fusion neutron source. Besides the possibility for steady state operation for a year or more, the geometry is chosen to avoid high loads on materials and plasma facing components. A comparatively small fusion hybrid device with "semi-poor" plasma confinement (with a low fusion Q factor) may be developed for industrial transmutation and energy production from spent nuclear fuel. This opportunity arises from a large fission to fusion energy multiplication ratio, Qr = Pfis/Pfus>>1. The upper bound on Qr is primarily determined by geometry and reactor safety. For the SFLM, the upper bound is Qr≈150, corresponding to a neutron multiplicity of keff=0.97. Power production in a mirror hybrid is predicted for a substantially lower electron temperature than the requirement Te≈10 keV for a fusion reactor. Power production in the SFLM seems possible with Q≈0.15, which is 10 times lower than typically anticipated for hybrids (and 100 times smaller than required for a fusion reactor). This relaxes plasma confinement demands, and broadens the range for use of plasmas with supra-thermal ions in hybrid reactors. The SFLM concept is based on a mirror machine stabilized by qudrupolar magnetic fields and large expander tanks beyond the confinement region. The purpose of the expander tanks is to distribute axial plasma loss flow over a sufficiently large area so that the receiving plates can withstand the heat. Plasma stability is not relying on a plasma flow into the expander regions. With a suppressed plasma flow into the expander tanks, a possibility arise for higher electron temperature. A brief presentation will be given on basic theory for the SFLM with plasma stability and electron temperature issues, RF heating computations with sloshing ion formation, neutron transport computations with reactor safety margins and material load estimates, magnetic coil designs as well as a discussion on

  16. 77 FR 16098 - In the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II...

    Science.gov (United States)

    2012-03-19

    ... the Matter of All Operating Boiling Water Reactor Licensees With Mark I and Mark II Containments... operate boiling-water reactors (BWRs) with Mark I and Mark II containment designs. II On March 11, 2011, a... Nuclear Reactor Regulation. Operating Boiling Water Reactor Licenses With Mark I and Mark II Containments...

  17. Light Water Reactor Sustainability Constellation Pilot Project FY11 Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    R. Johansen

    2011-09-01

    Summary report for Fiscal Year 2011 activities associated with the Constellation Pilot Project. The project is a joint effor between Constellation Nuclear Energy Group (CENG), EPRI, and the DOE Light Water Reactor Sustainability Program. The project utilizes two CENG reactor stations: R.E. Ginna and Nine Point Unit 1. Included in the report are activities associate with reactor internals and concrete containments.

  18. Continuous-flow solar UVB disinfection reactor for drinking water.

    Science.gov (United States)

    Mbonimpa, Eric Gentil; Vadheim, Bryan; Blatchley, Ernest R

    2012-05-01

    Access to safe, reliable sources of drinking water is a long-standing problem among people in developing countries. Sustainable solutions to these problems often involve point-of-use or community-scale water treatment systems that rely on locally-available resources and expertise. This philosophy was used in the development of a continuous-flow, solar UVB disinfection system. Numerical modeling of solar UVB spectral irradiance was used to define temporal variations in spectral irradiance at several geographically-distinct locations. The results of these simulations indicated that a solar UVB system would benefit from incorporation of a device to amplify ambient UVB fluence rate. A compound parabolic collector (CPC) was selected for this purpose. Design of the CPC was based on numerical simulations that accounted for the shape of the collector and reflectance. Based on these simulations, a prototype CPC was constructed using materials that would be available and inexpensive in many developing countries. A UVB-transparent pipe was positioned in the focal area of the CPC; water was pumped through the pipe to allow exposure of waterborne microbes to germicidal solar UVB radiation. The system was demonstrated to be effective for inactivation of Escherichia coli, and DNA-weighted UV dose was shown to govern reactor performance. The design of the reactor is expected to scale linearly, and improvements in process performance (relative to results from the prototype) can be expected by use of larger CPC geometry, inclusion of better reflective materials, and application in areas with greater ambient solar UV spectral irradiance than the location of the prototype tests. The system is expected to have application for water treatment among communities in (developing) countries in near-equatorial and tropical locations. It may also have application for disaster relief or military field operations, as well as in water treatment in areas of developed countries that receive

  19. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Paul Y [Los Alamos National Laboratory

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  20. Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

    OpenAIRE

    2009-01-01

    A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (S...

  1. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  2. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  3. Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, David William, E-mail: hummeld@mcmaster.ca; Novog, David Raymond

    2016-03-15

    Highlights: • A coupled spatial kinetics and thermalhydraulics model of the PT-SCWR was created. • Positive power excursions were demonstrated during accident-like transients. • The reactor will inherently self-shutdown in such transients with some delay. • A fast-acting shutdown system would limit the consequences of the power pulse. - Abstract: The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive

  4. Reactor vessel water level estimation during severe accidents using cascaded fuzzy neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Yeong; Yoo, Kwae Hwan; Choi, Geon Pil; Back, Ju Hyun; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-06-15

    Global concern and interest in the safety of nuclear power plants have increased considerably since the Fukushima accident. In the event of a severe accident, the reactor vessel water level cannot be measured. The reactor vessel water level has a direct impact on confirming the safety of reactor core cooling. However, in the event of a severe accident, it may be possible to estimate the reactor vessel water level by employing other information. The cascaded fuzzy neural network (CFNN) model can be used to estimate the reactor vessel water level through the process of repeatedly adding fuzzy neural networks. The developed CFNN model was found to be sufficiently accurate for estimating the reactor vessel water level when the sensor performance had deteriorated. Therefore, the developed CFNN model can help provide effective information to operators in the event of a severe accident.

  5. Hybrid reactor based on combined cavitation and ozonation: from concept to practical reality.

    Science.gov (United States)

    Gogate, P R; Mededovic-Thagard, S; McGuire, D; Chapas, G; Blackmon, J; Cathey, R

    2014-03-01

    The present work gives an in depth discussion related to the development of a hybrid advanced oxidation reactor, which can be effectively used for the treatment of various types of water. The reactor is based on the principle of intensifying degradation/disinfection using a combination of hydrodynamic cavitation, acoustic cavitation, ozone injection and electrochemical oxidation/precipitation. Theoretical studies have been presented to highlight the uniform distribution of the cavitational activity and enhanced generation of hydroxyl radicals in the cavitation zone, as well as higher turbulence in the main reactor zone. The combination of these different oxidation technologies have been shown to result in enhanced water treatment ability, which can be attributed to the enhanced generation of hydroxyl radicals, enhanced contact of ozone and contaminants, and the elimination of mass transfer resistances during electrochemical oxidation/precipitation. Compared to the use of individual approaches, the hybrid reactor is expected to intensify the treatment process by 5-20 times, depending on the application in question, which can be confirmed based on the literature illustrations. Also, the use of Ozonix® has been successfully proven while processing recycled fluids at commercial sites on over 750 oil and natural gas wells during hydraulic operations around the United States. The superiority of the hybrid process over conventional chemical treatments in terms of bacteria and scale reduction as well as increased water flowability and better chemical compatibility, which is a key requirement for oil and gas applications, has been established. Copyright © 2013 Elsevier B.V. All rights reserved.

  6. New instrumentation of reactor water level for PWR; Nueva Instrumentacion de nivel de agua del reactor para PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kaercher, S.

    2005-07-01

    Today, many PWR reactors are equipped with a reactor water level instrumentation system based on different measurement methods. Due to obsolescence issues, FRAMATOME ANP started to develop and quality a new water level measurement system using heated und unheated thermocouple measurements. the measuring principle is based on the fact that the heat transfer in water is considerably higher than in steam. The electronic cabinet for signal processing is based on a proven technology already developed, qualified and installed by FRAMATOME ANP in several NPPs. It is equipped with and advanced temperature measuring transducer for acquisition and processing of thermocouple signals. (Author)

  7. Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu/U and Th/U mixed fuels

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up, a tight pitch lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors. It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs. Various techniques were proposed to solve these problems. In this work, a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated. By utilizing numerical simulation technique, it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio (0.98), long burn-up (60 GWD/t) and negative void reactivity coefficients.

  8. Tritium distribution modeling in a Light Water New Production Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jaeckle, J.W.

    1989-05-01

    The tritium distribution and tritium release pathways in a new light water production reactor were examined. A computer model was developed to track the tritium as it makes its way through the various plant systems and ends up either as a release to the atmosphere, the cooling tower blowdown or to the solid waste system. The model was designed to predict the integrated yearly tritium releases and provide estimated airborne tritium concentrations in various locations within the plant. WNP-1 was used as a representative model for a Light Water New Production Reactor (LWNPR). The Tritium Distribution Model solves for the time dependent tritium concentration in a system of nodes. These nodes are connected to one another via a set of internodal flow paths and to various sources and sinks. For example, plant systems such as the primary system are the nodes, piping and leaks are the internodal flow paths, make-up water is a source, and release to the atmosphere is a sink. The expected water mass of each node; the flow rates between nodes, sources, and sinks; and tritium source rates are provided as input. The code will solve for the time dependent tritium concentration in each node and the amount of tritium ''released'' to the sinks. Preliminary calculations have been performed using WNP-1 plant specific information obtained primarily from the WNP-1 FSAR. Further work is currently in progress to refine the model and provide a more realistic set of input values which will better represent an operating LWNPR. 1 ref., 1 fig., 1 tab.

  9. Continuous supercritical water gasification of isooctane: A promising reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Susanti, Ratna F.; Kim, Jae-Duck; Kim, Jaehoon [Supercritical Fluid Research Laboratory, Clean Energy Center, Energy Division, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seoungbuk-gu, Seoul 136-791 (Korea); Department of Green Process and System Engineering, University of Science and Technology (UST), 113 Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea); Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Energy Center, Energy Division, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seoungbuk-gu, Seoul 136-791 (Korea); Lee, Youn-Woo [School of Chemical and Biological Engineering, Seoul National University, Gwanangro 599, Gwanak-gu, Seoul 151-744 (Korea)

    2010-03-15

    A new design of supercritical water gasification system was developed to achieve high hydrogen gas yield and good gas-liquid flow stability. The apparatus consisted of a reaction zone, an insulation zone and a cooling zone that were directly connected to the reaction zone. The reactor was set up at an inclination of 75 from vertical position, and feed and water were introduced at the bottom of the reactor. The performances of this new system were investigated with gasification of isooctane at various experimental conditions - reaction temperatures of 601-676 C, residence times of 6-33 s, isooctane concentrations of 5-33 wt%, and oxidant (hydrogen peroxide) concentrations up to 4507 mmol/L without using catalysts. A significant increase in hydrogen gas yield, almost four times higher than that from the previous up-down gasifier configuration (B. Veriansyah, J. Kim, J.D. Kim, Y.W. Lee, Hydrogen Production by Gasification of Isooctane using Supercritical Water, Int. J. Green Energy. 5 (2008) 322-333) was observed with the present gasifier configuration. High hydrogen gas yield (6.13 mol/mol isooctane) was obtained at high reaction temperature of 637 C, a low feed concentration of 9.9 wt% and a long residence time of 18 s in the presence of 2701.1 mmol/L hydrogen peroxide. At this condition, the produced gases mainly consisted of hydrogen (59.5 mol%), methane (14.8 mol%) and carbon dioxide (22.0 mol%), and a small amount of carbon monoxide (1.6 mol%) and C{sub 2}-C{sub 3} species (2.1 mol%). Reaction mechanisms of supercritical water gasification of isooctane were also presented. (author)

  10. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics

    Energy Technology Data Exchange (ETDEWEB)

    Brad Merrill; Melissa Teague; Robert Youngblood; Larry Ott; Kevin Robb; Michael Todosow; Chris Stanek; Mitchell Farmer; Michael Billone; Robert Montgomery; Nicholas Brown; Shannon Bragg-Sitton

    2014-02-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the United States’ nuclear industry. As a result, continual improvement of technology, including advanced materials and nuclear fuels, remains central to industry’s success. Decades of research combined with continual operation have produced steady advancements in technology and yielded an extensive base of data, experience, and knowledge on light water reactor (LWR) fuel performance under both normal and accident conditions. In 2011, following the Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex, enhancing the accident tolerance of LWRs became a topic of serious discussion. As a result of direction from the U.S. Congress, the U.S. Department of Energy Office of Nuclear Energy (DOE-NE) initiated an Accident Tolerant Fuel (ATF) Development program. The complex multiphysics behavior of LWR nuclear fuel makes defining specific material or design improvements difficult; as such, establishing qualitative attributes is critical to guide the design and development of fuels and cladding with enhanced accident tolerance. This report summarizes a common set of technical evaluation metrics to aid in the optimization and down selection of candidate designs. As used herein, “metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. Furthermore, this report describes a proposed technical evaluation methodology that can be applied to assess the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed for lead test rod or lead test assembly

  11. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    Science.gov (United States)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  12. Work Domain Analysis Methodology for Development of Operational Concepts for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-05-01

    This report describes a methodology to conduct a Work Domain Analysis in preparation for the development of operational concepts for new plants. This method has been adapted from the classical method described in the literature in order to better deal with the uncertainty and incomplete information typical of first-of-a-kind designs. The report outlines the strategy for undertaking a Work Domain Analysis of a new nuclear power plant and the methods to be used in the development of the various phases of the analysis. Basic principles are described to the extent necessary to explain why and how the classical method was adapted to make it suitable as a tool for the preparation of operational concepts for a new nuclear power plant. Practical examples are provided of the systematic application of the method and the various presentation formats in the operational analysis of advanced reactors.

  13. Research on the HYLIFE liquid-first-wall concept for future laser-fusion reactors. Final report No. 5

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.

    1980-09-01

    It has been proposed to protect the structural walls of a future laser fusion reactor with a curtain or fluid-wall of liquid lithium jets. As part of the investigation of this concept, experiments have been performed on planar sheet water jets issuing vertically downward from slit nozzles. The nozzles were subjected to transverse forced harmonic excitation to simulate the vibrational environment of the laser fusion reactor, and experiments were run at both 1 atm and at lower ambient pressures. Linear temporal stability theory is shown to predict the onset of the unstable regime and the initial spatial growth rates quite well for the cases where the amplitudes of the nozzle vibration are not too large and the waveform is nearly sinusoidal. In addition, both the linear theory and a simplified trajectory theory are shown to predict the initial wave envelope amplitudes very well. For larger amplitude nozzle excitation, the waveform becomes highly nonlinear and non-sinusoidal and can resemble a sawtooth waveform in some cases; these latter experimental results can only be partially explained by existing theories at the present time.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  15. Revised accident source terms for light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Soffer, L. [Nuclear Regulatory Commission, Washington, DC (United States)

    1995-02-01

    This paper presents revised accident source terms for light-water reactors incorporating the severe accident research insights gained in this area over the last 15 years. Current LWR reactor accident source terms used for licensing date from 1962 and are contained in Regulatory Guides 1.3 and 1.4. These specify that 100% of the core inventory of noble gases and 25% of the iodine fission products are assumed to be instantaneously available for release from the containment. The chemical form of the iodine fission products is also assumed to be predominantly elemental iodine. These assumptions have strongly affected present nuclear air cleaning requirements by emphasizing rapid actuation of spray systems and filtration systems optimized to retain elemental iodine. A proposed revision of reactor accident source terms and some im implications for nuclear air cleaning requirements was presented at the 22nd DOE/NRC Nuclear Air Cleaning Conference. A draft report was issued by the NRC for comment in July 1992. Extensive comments were received, with the most significant comments involving (a) release fractions for both volatile and non-volatile species in the early in-vessel release phase, (b) gap release fractions of the noble gases, iodine and cesium, and (c) the timing and duration for the release phases. The final source term report is expected to be issued in late 1994. Although the revised source terms are intended primarily for future plants, current nuclear power plants may request use of revised accident source term insights as well in licensing. This paper emphasizes additional information obtained since the 22nd Conference, including studies on fission product removal mechanisms, results obtained from improved severe accident code calculations and resolution of major comments, and their impact upon the revised accident source terms. Revised accident source terms for both BWRS and PWRS are presented.

  16. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

    2013-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  17. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

    2012-01-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

  18. Light Water Reactor Sustainability Program Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-01

    Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

  19. Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  20. Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

    2015-06-01

    “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

  1. Pressurized-water reactor internals aging degradation study. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    Luk, K.H. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pins and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.

  2. Technologies for Upgrading Light Water Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Daniel S. Wendt; Piyush Sabharwall; Vivek Utgikar

    2013-07-01

    Nuclear energy could potentially be utilized in hybrid energy systems to produce synthetic fuels and feedstocks from indigenous carbon sources such as coal and biomass. First generation nuclear hybrid energy system (NHES) technology will most likely be based on conventional light water reactors (LWRs). However, these LWRs provide thermal energy at temperatures of approximately 300°C, while the desired temperatures for many chemical processes are much higher. In order to realize the benefits of nuclear hybrid energy systems with the current LWR reactor fleets, selection and development of a complimentary temperature upgrading technology is necessary. This paper provides an initial assessment of technologies that may be well suited toward LWR outlet temperature upgrading for powering elevated temperature industrial and chemical processes during periods of off-peak power demand. Chemical heat transformers (CHTs) are a technology with the potential to meet LWR temperature upgrading requirements for NHESs. CHTs utilize chemical heat of reaction to change the temperature at which selected heat sources supply or consume thermal energy. CHTs could directly utilize LWR heat output without intermediate mechanical or electrical power conversion operations and the associated thermodynamic losses. CHT thermal characteristics are determined by selection of the chemical working pair and operating conditions. This paper discusses the chemical working pairs applicable to LWR outlet temperature upgrading and the CHT operating conditions required for providing process heat in NHES applications.

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

    Energy Technology Data Exchange (ETDEWEB)

    Conley, D.A.; Edson, J.L.; Fineman, C.F. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-03-01

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

  4. A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

    2002-03-15

    Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

  5. Development of Novel Water-Gas Shift Membrane Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ho, W. S. Winston

    2004-12-29

    This report summarizes the objectives, technical barrier, approach, and accomplishments for the development of a novel water-gas-shift (WGS) membrane reactor for hydrogen enhancement and CO reduction. We have synthesized novel CO{sub 2}-selective membranes with high CO{sub 2} permeabilities and high CO{sub 2}/H{sub 2} and CO{sub 2}/CO selectivities by incorporating amino groups in polymer networks. We have also developed a one-dimensional non-isothermal model for the countercurrent WGS membrane reactor. The modeling results have shown that H{sub 2} enhancement (>99.6% H{sub 2} for the steam reforming of methane and >54% H{sub 2} for the autothermal reforming of gasoline with air on a dry basis) via CO{sub 2} removal and CO reduction to 10 ppm or lower are achievable for synthesis gases. With this model, we have elucidated the effects of system parameters, including CO{sub 2}/H{sub 2} selectivity, CO{sub 2} permeability, sweep/feed flow rate ratio, feed temperature, sweep temperature, feed pressure, catalyst activity, and feed CO concentration, on the membrane reactor performance. Based on the modeling study using the membrane data obtained, we showed the feasibility of achieving H{sub 2} enhancement via CO{sub 2} removal, CO reduction to {le} 10 ppm, and high H{sub 2} recovery. Using the membrane synthesized, we have obtained <10 ppm CO in the H{sub 2} product in WGS membrane reactor experiments. From the experiments, we verified the model developed. In addition, we removed CO{sub 2} from a syngas containing 17% CO{sub 2} to about 30 ppm. The CO{sub 2} removal data agreed well with the model developed. The syngas with about 0.1% CO{sub 2} and 1% CO was processed to convert the carbon oxides to methane via methanation to obtain <5 ppm CO in the H{sub 2} product.

  6. Improved structure and long-life blanket concepts for heliotron reactors

    Science.gov (United States)

    Sagara, A.; Imagawa, S.; Mitarai, O.; Dolan, T.; Tanaka, T.; Kubota, Y.; Yamazaki, K.; Watanabe, K. Y.; Mizuguchi, N.; Muroga, T.; Noda, N.; Kaneko, O.; Yamada, H.; Ohyabu, N.; Uda, T.; Komori, A.; Sudo, S.; Motojima, O.

    2005-04-01

    New design approaches are proposed for the LHD-type heliotron D-T demo-reactor FFHR2 to solve the key engineering issues of blanket space limitation and replacement difficulty. A major radius of over 14 m is selected to permit a blanket-shield thickness of about 1 m and to reduce the neutron wall loading and toroidal field, while achieving an acceptable cost of electricity. Two sets of optimization are successfully carried out. One is to reduce the magnetic hoop force on the helical coil support structures by adjustment of the helical winding coil pitch parameter and the poloidal coils design, which facilitates expansion of the maintenance ports. The other is a long-life blanket concept using carbon armour tiles that soften the neutron energy spectrum incident on the self-cooled flibe-reduced activation ferritic steel blanket. In this adaptation of the spectral-shifter and tritium breeder blanket (STB) concept a local tritium breeding ratio over 1.2 is feasible by optimized arrangement of the neutron multiplier Be in the carbon tiles, and the radiation shielding of the superconducting magnet coils is also significantly improved. Using constant cross sections of a helically winding shape, the 'screw coaster' concept is proposed to replace in-vessel components such as the STB armour tiles. The key R&D issues for developing the STB concept, such as radiation effects on carbon and enhanced heat transfer of Flibe, are elucidated.

  7. Detecting pin diversion from pressurized water reactors spent fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Young S.; Sitaraman, Shivakumar

    2017-01-10

    Detecting diversion of spent fuel from Pressurized Water Reactors (PWR) by determining possible diversion including the steps of providing a detector cluster containing gamma ray and neutron detectors, inserting the detector cluster containing the gamma ray and neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray and neutron radiation responses of the gamma ray and neutron detectors in the guide tube holes, processing the gamma ray and neutron radiation responses at the guide tube locations by normalizing them to the maximum value among each set of responses and taking the ratio of the gamma ray and neutron responses at the guide tube locations and normalizing the ratios to the maximum value among them and producing three signatures, gamma, neutron, and gamma-neutron ratio, based on these normalized values, and producing an output that consists of these signatures that can indicate possible diversion of the pins from the spent fuel assembly.

  8. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  9. Aging assessment of PWR (Pressurized Water Reactor) Auxiliary Feedwater Systems

    Energy Technology Data Exchange (ETDEWEB)

    Casada, D.A.

    1988-01-01

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

  10. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) requesting exemptions from certain security requirements in Title 10 of the Code Federal Regulations (10 CFR) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has...

  11. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor...

  12. Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-01

    This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

  13. Final Report on Isotope Ratio Techniques for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gerlach, David C.; Gesh, Christopher J.; Hurley, David E.; Mitchell, Mark R.; Meriwether, George H.; Reid, Bruce D.

    2009-07-01

    The Isotope Ratio Method (IRM) is a technique for estimating the energy or plutonium production in a fission reactor by measuring isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to the cumulative energy production with standard reactor modeling methods.

  14. Application of fully ceramic microencapsulated fuels in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, C.; George, N.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States); Godfrey, A.; Terrani, K.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2012-07-01

    This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

  15. Application of Fully Ceramic Microencapsulated Fuels in Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL

    2012-01-01

    This study aims to perform a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in Light Water Reactors (LWRs). In particular pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor. Using uranium-based fuel and transuranic (TRU) based fuel in TRistructural ISOtropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher physical density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the TRU based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, feasibility of core designs fully loaded with TRU FCM lattices was demonstrated using the NESTLE three-dimensional core simulator.

  16. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  17. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout caused by external flooding using the RISMC toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego; Smith, Curtis; Prescott, Steven; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impacts of these two factors on the safety of the plant, the Risk Informed Safety Margin Characterization project aims to provide insights to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This paper focuses on the impacts of power uprate on the safety margin of a boiling water reactor for a flooding induced station black-out event. Analysis is performed by using a combination of thermal-hydraulic codes and a stochastic analysis tool currently under development at the Idaho National Laboratory, i.e. RAVEN. We employed both classical statistical tools, i.e. Monte-Carlo, and more advanced machine learning based algorithms to perform uncertainty quantification in order to quantify changes in system performance and limitations as a consequence of power uprate. Results obtained give a detailed investigation of the issues associated with a plant power uprate including the effects of station black-out accident scenarios. We were able to quantify how the timing of specific events was impacted by a higher nominal reactor core power. Such safety insights can provide useful information to the decision makers to perform risk informed margins management.

  18. ADDRESSING WATER FOOTPRINT CONCEPT: A DEMONSTRABLE STRATEGY FOR PAPERMAKING INDUSTRY

    OpenAIRE

    Jing Shen; Xueren Qian

    2012-01-01

    Since the introduction of the water footprint concept in 2002, in the context of humankind’s ever-increasing awareness of the valuable global freshwater resources, it has received more and more attention. The application of this relatively new concept has been expected to provide ecological and environmental benefits. For the water-intensive papermaking industry, it seems that water footprint needs to be addressed. The water footprint of cellulosic paper can be divided into three components, ...

  19. TREAT Neutronics Analysis of Water-Loop Concept Accommodating LWR 9-rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Hill, Connie M.; Woolstenhulme, Nicolas E.; Parry, James R.; Bess, John D.; Housley, Gregory K.

    2016-09-01

    Abstract. Simulation of a variety of transient conditions has been successfully achieved in the Transient Reactor Test (TREAT) facility during operation between 1959 and 1994 to support characterization and safety analysis of nuclear fuels and materials. A majority of previously conducted tests were focused on supporting sodium-cooled fast reactor (SFR) designs. Experiments evolved in complexity. Simulation of thermal-hydraulic conditions expected to be encountered by fuels and materials in a reactor environment was realized in the development of TREAT sodium loop experiment vehicles. These loops accommodated up to 7-pin fuel bundles and served to simulate more closely the reactor environment while safely delivering large quantities of energy into the test specimen. Some of the immediate TREAT restart operations will be focused on testing light water reactor (LWR) accident tolerant fuels (ATF). Similar to the sodium loop objectives, a water loop concept, developed and analyzed in the 1990’s, aimed at achieving thermal-hydraulic conditions encountered in commercial power reactors. The historic water loop concept has been analyzed in the context of a reactivity insertion accident (RIA) simulation for high burnup LWR 2-pin and 3-pin fuel bundles. Findings showed sufficient energy could be deposited into the specimens for evaluation. Similar results of experimental feasibility for the water loop concept (past and present) have recently been obtained using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. The old water loop concept required only two central TREAT core grid spaces. Preparation for future experiments has resulted in a modified water loop conceptual design designated the TREAT water environment recirculating loop (TWERL). The current TWERL design requires nine TREAT core grid spaces in order to place the water recirculating pump under the TREAT core. Due to the effectiveness of water moderation, neutronics analysis shows that removal of seven additional

  20. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    Directory of Open Access Journals (Sweden)

    Hyun-Gil Kim

    2016-02-01

    Full Text Available For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density composite pellet concepts are being developed as ATF pellets. A microcell UO2 pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts—surface-modified Zr-based alloy and SiC composite material—are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.

  1. Study of plutonium disposition using existing GE advanced Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1994-06-01

    The end of the cold war and the resulting dismantlement of nuclear weapons has resulted in the need for the US to dispose of 50 to 100 metric tons of excess of plutonium in a safe and proliferation resistant manner. A number of studies, including the recently released National Academy of Sciences (NAS) study, have recommended conversion of plutonium into spent nuclear fuel with its high radiation barrier as the best means of providing permanent conversion and long-term diversion resistance to this material. The NAS study ``Management and Disposition of Excess Weapons Plutonium identified Light Water Reactor spent fuel as the most readily achievable and proven form for the disposition of excess weapons plutonium. The study also stressed the need for a US disposition program which would enhance the prospects for a timely reciprocal program agreement with Russia. This summary provides the key findings of a GE study where plutonium is converted into Mixed Oxide (MOX) fuel and a typical 1155 MWe GE Boiling Water Reactor (BWR) is utilized to convert the plutonium to spent fuel. A companion study of the Advanced BWR has recently been submitted. The MOX core design work that was conducted for the ABWR enabled GE to apply comparable fuel design concepts and consequently achieve full MOX core loading which optimize plutonium throughput for existing BWRs.

  2. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Carson, S.D. [Sandia National Laboratories, New Mexico, NM (United States); Peterson, P.K. [Sandia National Laboratories, New Mexico, NM (United States)

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  3. Safety-Related Optimization and Analyses of an Innovative Fast Reactor Concept

    Directory of Open Access Journals (Sweden)

    Dalin Zhang

    2012-06-01

    Full Text Available Since a fast reactor core with uranium-plutonium fuel is not in its most reactive configuration under operating conditions, redistribution of the core materials (fuel, steel, sodium during a core disruptive accident (CDA may lead to recriticalities and as a consequence to severe nuclear power excursions. The prevention, or at least the mitigation, of core disruption is therefore of the utmost importance. In the current paper, we analyze an innovative fast reactor concept developed within the CP-ESFR European project, focusing on the phenomena affecting the initiation and the transition phases of an unprotected loss of flow (ULOF accident. Key phenomena for the initiation phase are coolant boiling onset and further voiding of the core that lead to a reactivity increase in the case of a positive void reactivity effect. Therefore, the first level of optimization involves the reduction, by design, of the positive void effect in order to avoid entering a severe accident. If the core disruption cannot be avoided, the accident enters into the transition phase, characterized by the progression of core melting and recriticalities due to fuel compaction. Dedicated features that enhance and guarantee a sufficient and timely fuel discharge are considered for the optimization of this phase.

  4. Science Concepts Young Children Learn through Water Play

    Science.gov (United States)

    Gross, Carol M.

    2012-01-01

    Water is fascinating, fun, and multifaceted. Children can play with it endlessly. But play, for play's sake, is not water's only value (Crosser, 1994, Tovey, 1993). Indeed, water play is a compelling focus of study for young children (Chalufour & Worth, 2005). The concepts that young children learn from water play are essential for early childhood…

  5. Science Concepts Young Children Learn through Water Play

    Science.gov (United States)

    Gross, Carol M.

    2012-01-01

    Water is fascinating, fun, and multifaceted. Children can play with it endlessly. But play, for play's sake, is not water's only value (Crosser, 1994, Tovey, 1993). Indeed, water play is a compelling focus of study for young children (Chalufour & Worth, 2005). The concepts that young children learn from water play are essential for early childhood…

  6. Light Water Reactor Sustainability Program: Integrated Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2017-05-01

    proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

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

  8. The water value-flow concept

    NARCIS (Netherlands)

    Seyam, I.M.; Hoekstra, A.Y.; Savenije, H.H.G.

    2003-01-01

    The value of water is a key issue in managing water resources in an efficient, equitable and sustainable way. Efforts to assess the value of water are often not linked to the properties of the natural water system, which makes it difficult to analyse upstream–downstream dependency. In order to accou

  9. The water value-flow concept

    NARCIS (Netherlands)

    Seyam, I.M.; Hoekstra, Arjen Ysbert; Savenije, H.H.G.

    2003-01-01

    The value of water is a key issue in managing water resources in an efficient, equitable and sustainable way. Efforts to assess the value of water are often not linked to the properties of the natural water system, which makes it difficult to analyse upstream–downstream dependency. In order to accou

  10. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide...

  11. Synergistic Water-Treatment Reactors Using a TiO2-Modified Ti-Mesh Filter

    Directory of Open Access Journals (Sweden)

    Akira Fujishima

    2013-07-01

    Full Text Available The recent applications of a TiO2-modified Ti-mesh filter (TMiP™ for water purification are summarized with newly collected data including biological assays as well as sewage water treatment. The water purification reactors consist of the combination of a TMiP, a UV lamp, an excimer VUV lamp, and an ozonation unit. The water purification abilities of the reactor were evaluated by decomposition of organic contaminants, inactivation of waterborne pathogens, and treatment efficiency for sewage water. The UV-C/TMiP/O3 reactor disinfected E. coli in aqueous suspension in approximately 1 min completely, and also decreased the number of E. coli in sewage water in 15 min dramatically. The observed rate constants of 7.5 L/min and 1.3 L/min were calculated by pseudo-first-order kinetic analysis respectively. Although organic substances in sewage water were supposed to prevent the UV-C/TMiP/O3 reactor from purifying water, the reactor reduced E. coli in sewage water continuously. On the other hand, although much higher efficiencies for decomposition of organic pollutants in water were achieved in the excimer/TMiP reactor, the disinfection activity of the reactor for waterborne pathogens was not as effective as the other reactors. The difference of efficiency between organic pollutants and waterborne pathogens in the excimer/TMiP reactor may be due to the size, the structure, and the decomposition mechanism of the organic pollutants and waterborne pathogens. These results show that a suitable system assisted by synergy of photocatalysts and other technologies such as ozonation has a huge potential as a practical wastewater purification system.

  12. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

  14. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  15. Grey water treatment in upflow anaerobic sludge blanket (UASB) reactor at different temperatures.

    Science.gov (United States)

    Elmitwalli, Tarek; Otterpohl, Ralf

    2011-01-01

    The treatment of grey water in two upflow anaerobic sludge blanket (UASB) reactors, operated at different hydraulic retention times (HRTs) and temperatures, was investigated. The first reactor (UASB-A) was operated at ambient temperature (14-25 degrees C) and HRT of 20, 12 and 8 h, while the second reactor (UASB-30) was operated at controlled temperature of 30 degrees C and HRT of 16, 10 and 6 h. The two reactors were fed with grey water from 'Flintenbreite' settlement in Luebeck, Germany. When the grey water was treated in the UASB reactor at 30 degrees C, total chemical oxygen demand (CODt) removal of 52-64% was achieved at HRT between 6 and 16 h, while at lower temperature lower removal (31-41%) was obtained at HRT between 8 and 20 h. Total nitrogen and phosphorous removal in the UASB reactors were limited (22-36 and 10-24%, respectively) at all operational conditions. The results showed that at increasing temperature or decreasing HRT of the reactors, maximum specific methanogenic activity of the sludge in the reactors improved. As the UASB reactor showed a significantly higher COD removal (31-64%) than the septic tank (11-14%) even at low temperature, it is recommended to use UASB reactor instead of septic tank (the most common system) for grey water pre-treatment. Based on the achieved results and due to high peak flow factor, a HRT between 8 and 12 h can be considered the suitable HRT for the UASB reactor treating grey water at temperature 20-30 degrees C, while a HRT of 12-24 h can be applied at temperature lower than 20 degrees C.

  16. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  17. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  18. Preliminary Safeguards Assessment for the Pebble-Bed Fluoride High-Temperature Reactor (PB-FHR) Concept

    Energy Technology Data Exchange (ETDEWEB)

    Disser, Jay; Arthur, Edward; Lambert, Janine

    2016-09-01

    This report examines a preliminary design for a pebble bed fluoride salt-cooled high temperature reactor (PB-FHR) concept, assessing it from an international safeguards perspective. Safeguards features are defined, in a preliminary fashion, and suggestions are made for addressing further nuclear materials accountancy needs.

  19. The use of LBB concept in French fast reactors: Application to SPX plant

    Energy Technology Data Exchange (ETDEWEB)

    Turbat, A.; Deschanels, H.; Sperandio, M. [and others

    1997-04-01

    The leak before break (LBB) concept was not used at the design level for SUPERPHENIX (SPX), but different studies have been performed or are in progress concerning different components : Main Vessel (MV), pipings. These studies were undertaken to improve the defense in depth, an approach used in all French reactors. In a first study, the LBB approach has been applied to the MV of SPX plant to verify the absence of risk as regards the core supporting function and to help in the definition of in-service inspection (ISI) program. Defining a reference semi-elliptic defect located in the welds of the structure, it is verified that the crack growth is limited and that the end-of-life defect is smaller than the critical one. Then it is shown that the hoop welds (those which are the most important for safety) located between the roof and the triple point verify the leak-before-break criteria. However, generally speaking, the low level of membrane primary stresses which is favorable for the integrity of the vessel makes the application of the leak-before-break concept more difficult due to small crack opening areas. Finally, the extension of the methodology to the secondary pipings of SPX incorporating recent European works of DCRC is briefly presented.

  20. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  1. Advanced Fuels Campaign Light Water Reactor Accident Tolerant Fuel Performance Metrics Executive Summary

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton

    2014-02-01

    Research and development (R&D) activities on advanced, higher performance Light Water Reactor (LWR) fuels have been ongoing for the last few years. Following the unfortunate March 2011 events at the Fukushima Nuclear Power Plant in Japan, the R&D shifted toward enhancing the accident tolerance of LWRs. Qualitative attributes for fuels with enhanced accident tolerance, such as improved reaction kinetics with steam resulting in slower hydrogen generation rate, provide guidance for the design and development of fuels and cladding with enhanced accident tolerance. A common set of technical metrics should be established to aid in the optimization and down selection of candidate designs on a more quantitative basis. “Metrics” describe a set of technical bases by which multiple concepts can be fairly evaluated against a common baseline and against one another. This report describes a proposed technical evaluation methodology that can be applied to evaluate the ability of each concept to meet performance and safety goals relative to the current UO2 – zirconium alloy system and relative to one another. The resultant ranked evaluation can then inform concept down-selection, such that the most promising accident tolerant fuel design option(s) can continue to be developed toward qualification.

  2. Research and Development Methodology for Practical Use of Accident Tolerant Fuel in Light Water Reactors

    Directory of Open Access Journals (Sweden)

    Masaki Kurata

    2016-02-01

    Full Text Available Research and development (R&D methodology for the practical use of accident tolerant fuel (ATF in commercial light water reactors is discussed in the present review. The identification and quantification of the R&D-metrics and the attribute of candidate ATF-concepts, recognition of the gap between the present R&D status and the targeted practical use, prioritization of the R&D, and technology screening schemes are important for achieving a common understanding on technology screening process among stakeholders in the near term and in developing an efficient R&D track toward practical use. Technology readiness levels and attribute guides are considered to be proper indices for these evaluations. In the midterm, the selected ATF-concepts will be developed toward the technology readiness level-5, at which stage the performance of the prototype fuel rods and the practicality of industrial scale fuel manufacturing will be verified and validated. Regarding the screened-out concepts, which are recognized to have attractive potentials, the fundamental R&D should be continued in the midterm to find ways of addressing showstoppers.

  3. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    Energy Technology Data Exchange (ETDEWEB)

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

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

  5. Bacterial Colonization of Pellet Softening Reactors Used during Drinking Water Treatment

    NARCIS (Netherlands)

    Hammes, F.; Boon, N.; Vital, M.; Ross, P.; Magic-Knezev, A.; Dignum, M.

    2010-01-01

    Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase t

  6. 77 FR 23513 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-04-19

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... Management Criteria for PWR Reactor Vessel Internal Components.'' The original notice provided the ADAMS... published a notice requesting public comments on draft LR-ISG-2011-04, ``Updated Aging Management...

  7. Water Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    Abuaf, N.; Levine, M.M.; Saha, P.; van Rooyen, D.

    1980-08-01

    The Water Reactor Safety Research Programs quarterly report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: LWR Thermal Hydraulic Development, Advanced Code Evlauation, TRAC Code Assessment, and Stress Corrosion Cracking of PWR Steam Generator Tubing.

  8. Impact of inflow transport approximation on light water reactor analysis

    Science.gov (United States)

    Choi, Sooyoung; Smith, Kord; Lee, Hyun Chul; Lee, Deokjung

    2015-10-01

    The impact of the inflow transport approximation on light water reactor analysis is investigated, and it is verified that the inflow transport approximation significantly improves the accuracy of the transport and transport/diffusion solutions. A methodology for an inflow transport approximation is implemented in order to generate an accurate transport cross section. The inflow transport approximation is compared to the conventional methods, which are the consistent-PN and the outflow transport approximations. The three transport approximations are implemented in the lattice physics code STREAM, and verification is performed for various verification problems in order to investigate their effects and accuracy. From the verification, it is noted that the consistent-PN and the outflow transport approximations cause significant error in calculating the eigenvalue and the power distribution. The inflow transport approximation shows very accurate and precise results for the verification problems. The inflow transport approximation shows significant improvements not only for the high leakage problem but also for practical large core problem analyses.

  9. Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

    2006-01-15

    This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-01-16

    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.

  11. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

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

  12. Basic Concepts of Magnetic Water Treatment

    Directory of Open Access Journals (Sweden)

    Oleg Mosin

    2014-06-01

    Full Text Available This review article outlines an overview of new trends and modern approaches for practical implementation of magnetic water treatment to eliminate scaling salts (carbonate, chloride and sulfate salts of Ca2+, Mg2+, Fe2+ and Fe3+ cations in power heat-exchanger devices and pipe lines. The principles of physical effects of the magnetic field on H2O molecules as well as the parameters of physico-chemical processes occurring in water and the behavior of the dissolved in water scaling salts subjected to the magnetic treatment are discussed. It is demonstrated that the effect of the magnetic field on water is a complex multifactorial phenomenon resulted in changes of the structure of hydrated ions as well as the physico-chemical properties and behavior of dissolved inorganic salts, changes in the rate of electrochemical coagulation and aggregate stability (clumping and consolidation, formation of multiple nucleation sites on the particles of fine dispersed precipitate consisting of crystals of substantially uniform size. There are also submitted data on constructive features of various magnetic water treatment devices produced by domestic industry, based on the permanent magnets and electromagnets (solenoids, such as hydro magnetic systems (HMS, magnetic transducers (MT and magnetic activators (MA of water. It was estimated the efficiency of using the various magnetic water treatment devices in water treatment technologies.

  13. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-09-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

  14. Ferritic Alloys as Accident Tolerant Fuel Cladding Material for Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rebak, Raul B. [General Electric Global Research, Schnectady, NY (United States)

    2014-12-30

    The objective of the GE project is to demonstrate that advanced steels such as iron-chromium-aluminum (FeCrAl) alloys could be used as accident tolerant fuel cladding material in commercial light water reactors. The GE project does not include fuel development. Current findings support the concept that a FeCrAl alloy could be used for the cladding of commercial nuclear fuel. The use of this alloy will benefit the public since it is going to make the power generating light water reactors safer. In the Phase 1A of this cost shared project, GE (GRC + GNF) teamed with the University of Michigan, Los Alamos National Laboratory, Brookhaven National Laboratory, Idaho National Laboratory, and Oak Ridge National Laboratory to study the environmental and mechanical behavior of more than eight candidate cladding materials both under normal operation conditions of commercial nuclear reactors and under accident conditions in superheated steam (loss of coolant condition). The main findings are as follows: (1) Under normal operation conditions the candidate alloys (e.g. APMT, Alloy 33) showed excellent resistance to general corrosion, shadow corrosion and to environmentally assisted cracking. APMT also showed resistance to proton irradiation up to 5 dpa. (2) Under accident conditions the selected candidate materials showed several orders of magnitude improvement in the reaction with superheated steam as compared with the current zirconium based alloys. (3) Tube fabrication feasibility studies of FeCrAl alloys are underway. The aim is to obtain a wall thickness that is below 400 µm. (4) A strategy is outlined for the regulatory path approval and for the insertion of a lead fuel assembly in a commercial reactor by 2022. (5) The GE team worked closely with INL to have four rodlets tested in the ATR. GE provided the raw stock for the alloys, the fuel for the rodlets and the cost for fabrication/welding of the rodlets. INL fabricated the rodlets and the caps and welded them to

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  16. Conception of high safety reactor MAVR, technical and economical fuel cycle characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.M.; Cherepnin, Yu.S.

    1993-12-31

    Operation safety of reactor MAVR under nominal and emergency situations is based on creation of conditions for the minimum time of fuel operation in the core at the minimum quantity of the fissionable material. The variants of core elements construction, of the reactor control systems, and the possible scheme of fuel cycles of the reactor MAVR are considered.

  17. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

  19. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos, E-mail: lcastro@instec.cu, E-mail: leored1984@gmail.com, E-mail: agamezgmf@gmail.com, E-mail: jrosales@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Oliveira, Carlos Brayner de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Pos-Graduacao em Modelagem Computacional

    2015-07-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  20. Light Water Reactor Sustainability Program: Analysis of Pressurized Water Reactor Station Blackout Caused by External Flooding Using the RISMC Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-08-01

    The existing fleet of nuclear power plants is in the process of extending its lifetime and increasing the power generated from these plants via power uprates. In order to evaluate the impact of these factors on the safety of the plant, the Risk Informed Safety Margin Characterization (RISMC) project aims to provide insight to decision makers through a series of simulations of the plant dynamics for different initial conditions (e.g., probabilistic analysis and uncertainty quantification). This report focuses, in particular, on the application of a RISMC detailed demonstration case study for an emergent issue using the RAVEN and RELAP-7 tools. This case study looks at the impact of a couple of challenges to a hypothetical pressurized water reactor, including: (1) a power uprate, (2) a potential loss of off-site power followed by the possible loss of all diesel generators (i.e., a station black-out event), (3) and earthquake induces station-blackout, and (4) a potential earthquake induced tsunami flood. The analysis is performed by using a set of codes: a thermal-hydraulic code (RELAP-7), a flooding simulation tool (NEUTRINO) and a stochastic analysis tool (RAVEN) – these are currently under development at the Idaho National Laboratory.

  1. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  2. Energy efficient electrocoagulation using a new flow column reactor to remove nitrate from drinking water - Experimental, statistical, and economic approach.

    Science.gov (United States)

    Hashim, Khalid S; Shaw, Andy; Al Khaddar, Rafid; Pedrola, Montserrat Ortoneda; Phipps, David

    2017-03-09

    In this investigation, a new bench-scale electrocoagulation reactor (FCER) has been applied for drinking water denitrification. FCER utilises the concepts of flow column to mix and aerate the water. The water being treated flows through the perforated aluminium disks electrodes, thereby efficiently mixing and aerating the water. As a result, FCER reduces the need for external stirring and aerating devices, which until now have been widely used in the electrocoagulation reactors. Therefore, FCER could be a promising cost-effective alternative to the traditional lab-scale EC reactors. A comprehensive study has been commenced to investigate the performance of the new reactor. This includes the application of FCER to remove nitrate from drinking water. Estimation of the produced amount of H2 gas and the yieldable energy from it, an estimation of its preliminary operating cost, and a SEM (scanning electron microscope) investigation of the influence of the EC process on the morphology of the surface of electrodes. Additionally, an empirical model was developed to reproduce the nitrate removal performance of the FCER. The results obtained indicated that the FCER reduced the nitrate concentration from 100 to 15 mg/L (World Health Organization limitations for infants) after 55 min of electrolysing at initial pH of 7, GBE of 5 mm, CD of 2 mA/cm(2), and at operating cost of 0.455 US $/m(3). Additionally, it was found that FCER emits H2 gas enough to generate a power of 1.36 kW/m(3). Statistically, the relationship between the operating parameters and nitrate removal could be modelled with R(2) of 0.848. The obtained SEM images showed a large number dents on anode's surface due to the production of aluminium hydroxides.

  3. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  4. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  5. Reactor physics and safety aspects of various design options of a Russian light water reactor with rock-like fuels

    Science.gov (United States)

    Bondarenko, A. V.; Komissarov, O. V.; Kozmenkov, Ya. K.; Matveev, Yu. V.; Orekhov, Yu. I.; Pivovarov, V. A.; Sharapov, V. N.

    2003-06-01

    This paper presents results of analytical studies on weapons grade plutonium incineration in VVER (640) medium size light water reactors using a special composition of rock-like fuel (ROX-fuel) to assure spent fuel long-term storage without its reprocessing. The main goal is to achieve high degree of plutonium incineration in once-through cycle. In this paper we considered two fuel compositions. In both compositions weapons grade plutonium is used as fissile material. Spinel (MgAl 2O 4) is used as the 'preserving' material assuring safe storage of the spent fuel. Besides an inert matrix, the option of rock-like fuel with thorium dioxide was studied. One of principal problems in the realization of the proposed approach is the substantial change of properties of the light water reactor core when passing to the use of the ROX-fuel, in particular: (i) due to the absence of 238U the Doppler effect playing a crucial role in reactor's self-regulation and limiting the consequences of reactivity accidents, decreases significantly, (ii) no fuel breeding on one hand, and the quest to attain the maximum plutonium burnup on the other hand, would result in a drastical change of the fuel assembly power during the lifetime and, as a consequence, the rise in irregularity of the power density of fuel assemblies, (iii) both the control rods worth and dissolved boron worth decrease in view of neutron spectrum hardening brought on by the larger absorption cross-section of plutonium as compared to uranium, (iv) βeff is markedly reduced. All these distinctive features are potentially detrimental to the reactor nuclear safety. The principal objective of this work is that to identify a variant of the fuel composition and the reactor layout, which would permit neutralize the negative effect of the above-mentioned distinctive features.

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

    Science.gov (United States)

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

    2016-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

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

  8. An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gentry, Cole A [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL; Powers, Jeffrey J [ORNL; Maldonado, G Ivan [ORNL

    2014-01-01

    An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

  9. Coupled anaerobic-aerobic treatment of whey wastewater in a sequencing batch reactor: proof of concept.

    Science.gov (United States)

    Frigon, J C; Bruneau, T; Moletta, R; Guiot, S R

    2007-01-01

    A proof of concept was performed in order to verify if the coupling of anaerobic and aerobic conditions inside the same digester could efficiently treat a reconstituted whey wastewater at 21 degrees C. The sequencing batch reactor (SBR) cycles combined initial anaerobic phase and final aerobic phase with reduced aeration. A series of 24 h cycles in 0.5 L digesters, with four different levels of oxygenation (none, 54, 108 and 182 mgO2 per gram of chemical oxygen demand (COD)), showed residual soluble chemical oxygen demand (sCOD) of 683 +/- 46, 720 +/- 33, 581 +/- 45, 1239 +/- 15 mg L(-1), respectively. Acetate and hydrogen specific activities were maintained for the anaerobic digester, but decreased by 10-25% for the acetate and by 20-50% for the hydrogen, in the coupled digesters. The experiment was repeated using 48 h cycles with limited aeration during 6 or 16 hours at 54 and 108 mgO2gCODinitial(-1), displaying residual sCOD of 177 +/- 43, 137 +/- 38, 104 +/- 22 and 112 +/- 9 mgL(-1) for the anaerobic and the coupled digesters, respectively. The coupled digesters recovered after a pH shock with residual sCOD as low as 132 mg L(-1) compared to 636 mg L(-1) for the anaerobic digester. With regard to the obtained results, the feasibility of the anaerobic-aerobic coupling in SBR digesters for the treatment of whey wastewater was demonstrated.

  10. FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

    1996-09-01

    The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

  11. Conceptual design and economic analysis of a light water reactor fuel enricher/regenerator. FY 1978 year-end report

    Energy Technology Data Exchange (ETDEWEB)

    Grand, P; Kouts, H J; Powell, J R; Steinberg, M; Takahashi, H

    1979-05-01

    A study has been performed to evaluate the use of high-energy particle accelerators as nuclear fuel enrichers and nuclear fuel regenerators. This builds on ideas that have been current for many years. The new study has, however, explored some novel approaches that have not been examined before. A specific conceptual system chosen for more detailed study would stretch the energy available from natural uranium by a factor of about 3, reduce the separative work requirements by a factor of about 4, and reduce the volume of spent fuel to be stored by a factor of 2, compared to the current once-through light water reactor (LWR) fuel cycle. The concept avoids the need for chemical reprocessing of spent fuel, and would permit continued use of LWR's beyond the time when limitations on fuel resources might otherwise lead to their being phased out. This concept, which is called the Linear Accelerator Fuel Enricher/Regenerator, is therefore viewed as offering a practical means of stretching the use of the nuclear fuel resource in the framework of the existing light water reactor fuel cycle. This report describes and analyzes the concept referred to. An explanation of the principles underlying the concept is given. Particular attention is devoted to engineering feasibility, proliferation resistance, and economics. It is seen that the concept draws on only proven technology as regards bothaccelerator design and the fuel irradiation process, and is adapted to existing LWR designs with no change except in fuel-handling practices. A preliminary evaluation of radiation damage, coolant options, and power conversion systems is provided. Neutronic, thermal-hydraulic, and burnup calculations are presented. An analysis is made of fuel economy. Approximate costs of electric power produced using this concept are evaluated and discussed. Estimated development costs of commercialization are provided.

  12. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  13. Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Nobuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Izumi, Fumio

    1995-12-01

    The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author).

  14. Performance of materials in the component cooling water systems of pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.S.

    1993-06-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed.

  15. Modelling of turbulent hydrocarbon combustion. Test of different reactor concepts for describing the interactions between turbulence and chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Kremer, H. [Ruhr-Universitaet Bochum, Lehrstuhl fuer Energieanlagentechnik, Bochum (Germany); Kilpinen, P.; Hupa, M. [Aabo Akademi, Turku (Finland). Combustion Chemistry Research Group

    1997-12-31

    The detailed modelling of turbulent reactive flows with CFD-codes is a major challenge in combustion science. One method of combining highly developed turbulence models and detailed chemistry in CFD-codes is the application of reactor based turbulence chemistry interaction models. In this work the influence of different reactor concepts on methane and NO{sub x} chemistry in turbulent reactive flows was investigated. Besides the classical reactor approaches, a plug flow reactor (PFR) and a perfectly stirred reactor (PSR), the Eddy-Dissipation Combustion Model (EDX) and the Eddy Dissipation Concept (EDC) were included. Based on a detailed reaction scheme and a simplified 2-step mechanism studies were performed in a simplified computational grid consisting of 5 cells. The investigations cover a temperature range from 1273 K to 1673 K and consider fuel-rich and fuel-lean gas mixtures as well as turbulent and highly turbulent flow conditions. All test cases investigated in this study showed a strong influence of the reactor residence time on the species conversion processes. Due to this characteristic strong deviations were found for the species trends resulting from the different reactor approaches. However, this influence was only concentrated on the `near burner region` and after 4-5 cells hardly any deviation and residence time dependence could be found. The importance of the residence time dependence increased when the species conversion was accelerated as it is the case for overstoichiometric combustion conditions and increased temperatures. The study focused furthermore on the fine structure in the EDC. Unlike the classical approach this part of the cell was modelled as a PFR instead of a PSR. For high temperature conditions there was hardly any difference between both reactor types. However, decreasing the temperature led to obvious deviations. Finally, the effect of the selective species transport between the cells on the conversion process was investigated

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

    Directory of Open Access Journals (Sweden)

    Mirmohammadi Sayedeh Leila

    2013-01-01

    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.

  17. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  18. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zdarek, J.; Pecinka, L. [Nuclear Research Institute Rez (Czech Republic)

    1997-04-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors.

  19. ``Virtual water'': An unfolding concept in integrated water resources management

    Science.gov (United States)

    Yang, Hong; Zehnder, Alexander

    2007-12-01

    In its broadest sense, virtual water refers to the water required for the production of food commodities. Issues relating to virtual water have drawn much attention in scientific communities and the political sphere since the mid 1990s. This paper provides a critical review of major research issues and results in the virtual water literature and pinpoints the remaining questions and the direction of research in future virtual water studies. We conclude that virtual water studies have helped to raise the awareness of water scarcity and its impact on food security and to improve the understanding of the role of food trade in compensating for water deficit. However, the studies so far have been overwhelmingly concerned with the international food trade, and many solely quantified virtual water flows associated with food trade. There is a general lack of direct policy relevance to the solutions to water scarcity and food insecurity, which are often local, regional, and river basin issues. The obscurity in the conceptual basis of virtual water also entails some confusion. The methodologies and databases of the studies are often crude, affecting the robustness and reliability of the results. Looking ahead, future virtual water studies need to enhance the policy relevance by strengthening their linkages with national and regional water resources management. Meanwhile, integrated approaches taking into consideration the spatial and temporal variations of blue and green water resources availability and the complexity of natural, socioeconomic, and political conditions are necessary in assessing the trade-offs of the virtual water strategy in dealing with water scarcity. To this end, interdisciplinary efforts and quantitative methods supported by improved data availability are greatly important.

  20. Modeling of Fischer-Tropsch Synthesis in a Slurry Reactor with Water Permeable Membrane

    Institute of Scientific and Technical Information of China (English)

    Fabiano A. N. Fernandes

    2007-01-01

    Fischer-Tropsch synthesis is an important chemical process for the production of liquid fuels and olefins. In recent years, the abundant availability of natural gas and the increasing demand of olefins, diesel, and waxes have led to a high interest to further develop this process. A mathematical model of a slurry membrane reactor used for syngas polymerization was developed to simulate and compare the maximum yields and operating conditions in the reactor with that in a conventional slurry reactor.The carbon polymerization was studied from a modeling point of view in a slurry reactor with a water permeable membrane and a conventional slurry reactor. Simulation results show that different parameters affect syngas conversion and carbon product distribution, such as the hydrogen to carbon monoxide ratio,and the membrane parameters such as membrane permeance.

  1. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  2. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  3. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  4. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    Energy Technology Data Exchange (ETDEWEB)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.

    2001-07-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 {approx} 10{sup -V} at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F.

    1984-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  7. Optimization and simplification of the concept of non-moderated Thorium Molten Salt Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Merle-Lucotte, Elsa; Heuer, Daniel; Allibert, Michel; Doligez, Xavier; Ghetta, Veronique; Le Brun, Christian [LPSC-IN2P3-CNRS/UJF/INPG, LPSC 53 avenue des Martyrs, 38026 Grenoble Cedex (France)

    2008-07-01

    Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria defined by the Generation IV forum, and may be operated in simplified and safe conditions in the Th/{sup 233}U fuel cycle with fluoride salts. The characteristics of the non-moderated TMSR based on a fast neutron spectrum are detailed in this paper: we aimed at designing an optimised TMSR with the simplest configuration. Using a simple kinetic-point model, we analyze the reactor's transient as the total reactivity margins are introduced in the core. We thus confirm, beyond the classical advantages of molten salt reactors, the satisfactory behaviour of the TMSR in terms of safety and the excellent level of stability which can be achieved in such reactors. (authors)

  8. Solar disinfection of contaminated water: a comparison of three small-scale reactors

    Energy Technology Data Exchange (ETDEWEB)

    McLoughlin, O.A.; Gill, L.W. [Dublin Univ. (Ireland). Dept. of Civil, Structural and Environmental Engineering; Kehoe, S.C. [Royal College of Surgeons in Ireland, Dublin (Ireland). Dept. of Surgery; McGuigan, K.G.; Duffy, E.F.; Al Touati, F. [Royal College of Surgeons in Ireland, Dublin (Ireland). Dept. of Physiology and Medical Physics; Gernjak, W.; Alberola, I.O.; Malato Rodriguez, S. [Plataforma Solar de Almeria (CIEMAT), Tabemas (Spain)

    2004-11-01

    This paper compares three different collector shapes for the disinfection of water heavily contaminated with Escherichia coli (K-12). Tests were carried out in real sunlight using laboratory scale reactors to determine the performance of different reflector profiles. The reactors were constructed using Pyrex tubing and aluminium reflectors of compound parabolic, parabolic and V-groove profiles. Results have shown that the compound parabolic reflector promoted a more successful inactivation of E. coli than the parabolic and V-groove profiles. Tests were also carried out to assess the improvement to disinfection which could be achieved using TiO{sub 2} coated Pyrex rods fixed within the reactors. This technique, however, yielded a slight enhancement in the compound parabolic reactor but no benefit to overall disinfection performance in either the parabolic or V-groove reactors. These results show that the use of UV sunlight to disinfect contaminated drinking water in a full-scale continuous flow solar reactor is both promising and an appropriate technology for developing countries but that the inclusion of a fixed photocatalyst within the reactor tubes has yet to prove any significant improvement. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-02-01

    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.

  10. Low-temperature water reactor for the district heating atomic power plant

    Energy Technology Data Exchange (ETDEWEB)

    Skvortsov, S.A.; Sokolov, I.N.; Krauze, L.V.; Nikiporetz, Yu.G.; Philimonov, Y.V.

    1978-04-01

    A natural convection low-pressure water reactor can be utilized as a source of district heating. This provides inherent safety factors under conditions requiring emergency core cooling. The reactor pressure vessel is contained within a prestressed concrete shell, both of which are designed to withstand accident overpressure. This also results in a relatively thin-walled reactor vessel that can be fabricated on-site. The overall safety and economy of such a system merits further consideration as a system for providing low-temperature nuclear heat for district heating.

  11. Water Recovery System Architecture and Operational Concepts to Accommodate Dormancy

    Science.gov (United States)

    Carter, Layne; Tabb, David; Anderson, Molly

    2017-01-01

    Future manned missions beyond low Earth orbit will include intermittent periods of extended dormancy. The mission requirement includes the capability for life support systems to support crew activity, followed by a dormant period of up to one year, and subsequently for the life support systems to come back online for additional crewed missions. NASA personnel are evaluating the architecture and operational concepts that will allow the Water Recovery System (WRS) to support such a mission. Dormancy could be a critical issue due to concerns with microbial growth or chemical degradation that might prevent water systems from operating properly when the crewed mission began. As such, it is critical that the water systems be designed to accommodate this dormant period. This paper identifies dormancy issues, concepts for updating the WRS architecture and operational concepts that will enable the WRS to support the dormancy requirement.

  12. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    OpenAIRE

    Urva Pancholi; Dhaval Dave; Ajay Patel

    2016-01-01

    The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR), generally of a Canadian Deuterium Uranium (CANDU) type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniqu...

  13. IRIS Reactor a Suitable Option to Provide Energy and Water Desalination for the Mexican Northwest Region

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, G.; Ramirez, R.; Gomez, C.; Viais, J.

    2004-10-03

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity. The IRIS reactor offers a very suitable source of energy given its modular size of 300 MWe and it can be coupled with a desalination plant to provide the potable water for human consumption, agriculture and industry. The present paper assess the water and energy requirements for the Northwest region of Mexico and how the deployment of the IRIS reactor can satisfy those necessities. The possible sites for deployment of Nuclear Reactors are considered given the seismic constraints and the closeness of the sea for external cooling. And in the other hand, the size of the desalination plant and the type of desalination process are assessed accordingly with the water deficit of the region.

  14. THE INFLUENCE OF MIEX® RESIN FOR WATER TREATMENT EFFICIENCYIN A HYBRID MEMBRANE REACTOR

    Directory of Open Access Journals (Sweden)

    Mariola Rajca

    2014-10-01

    Full Text Available The paper presents the results of studies related to the effectiveness of removal of natural organic matter (NOM from water using hybrid membrane reactor in which ion exchange and ultrafiltration processes were performed. MIEX® resin by Orica Watercare and immersed ultrafiltration polyvinylidene fluoride capillary module ZeeWeed 1 (ZW 1 by GE Power&Water operated at negative pressure were used. The application of multifunctional reactor had a positive effect on the removal of contaminants and enabled the production of high quality water. Additionally, in refer to single stage ultrafiltration it minimalized the occurrence of membrane fouling.

  15. Irradiation tests of texture controlled cladding for pressurized water reactor in foreign reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ukawa, Kazunori [Hokkaido Electric Power Co., Inc., Sapporo (Japan); Abeta, Sadaaki

    1996-12-01

    PWR electric power companies and makers are promoting a high burn-up program for nuclear fuel aiming at reducing spent fuel. PCI tolerance performance of fuel clad is desirable to promote a high burnup program. It is clear that when we change the direction of a Zircaloy crystal structure by improving a manufacture process (texture control), PCI tolerance performance greatly improved. In this study, an improved clad was burnt in a R2 reactor to 61.5 GWd/t and a power ramp test was carried out in Sweden. Based on irradiation data, power ramp test data and post irradiation examination data, improvement of PCI tolerance performance was confirmed. (author)

  16. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detector

    Energy Technology Data Exchange (ETDEWEB)

    Hellfeld, D. [Texas A & M Univ., College Station, TX (United States); Dazeley, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bernstein, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marianno, C. [Texas A & M Univ., College Station, TX (United States)

    2015-11-25

    The potential of elastic antineutrino-electron scattering (ν¯e + e → ν¯e + e) in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13 km standoff from a 3.758 GWt light water nuclear reactor. Background was estimated via independent simulations and by appropriately scaling published measurements from similar detectors. Many potential backgrounds were considered, including solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclide and water-borne radon decays, and gamma rays from the photomultiplier tubes, detector walls, and surrounding rock. The detector response was modeled using a GEANT4-based simulation package. The results indicate that with the use of low radioactivity PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. The directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. Lastly, the results provide a list of theoretical conditions that, if satisfied in practice, would enable nuclear reactor antineutrino directionality in a Gd-doped water Cherenkov detector approximately 10 km from a large power reactor.

  17. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    Science.gov (United States)

    Fletcher, C. D.

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes.

  18. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... 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...

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

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

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

  20. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  1. Inspection and evaluation guidelines for light water reactor internals

    Energy Technology Data Exchange (ETDEWEB)

    Iizuka, N. [Tokyo Electric Power Co., Inc. (Japan); Taniguchi, K. [Kansai Electric Power Co., Inc., Osaka (Japan); Yoshinaga, T. [Japan Atomic Power Co., Tokyo (Japan)

    2002-12-01

    On February, 2000, in the Engineering Society of Thermal and Nuclear Power Generation, the 'Investigation Group on Inspection and Evaluation Guidelines for Nuclear Reactor Internals' was established. This group was started at moments of some damage cases on reactor internals on BWRs and PWRs in Japan and foreign countries and of finding out cracks based on a number of SCC (stress corrosion cracking) at Inconel alloy weldings of a shroud support of BWR internals in Tsuruga Nuclear Power Station Unit-1 of the Japan Nuclear Power Generation Co., Ltd. on December, 1999. Under these conditions, this group made some guidelines for rational inspection with clear technical foundation, and so on as well as arrangements on structural functions, importance at safety, and so on of the reactor internals, promoted some investigations aiming at wide general proposal on how to carry out future internal inspections on LWR in Japan, and completed almost all of the investigations on March, 2002. Here were described basic indications of the guideline development and summaries of the developed guideline. (G.K.)

  2. Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

    Science.gov (United States)

    Martin, James J.; Reid, Robert S.

    2004-01-01

    A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

  3. A spherical torus nuclear fusion reactor space propulsion vehicle concept for fast interplanetary travel

    Science.gov (United States)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1999-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a>5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all major systems including payload, central truss, nuclear reactor (including diverter and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, and component design.

  4. A Spherical Torus Nuclear Fusion Reactor Space Propulsion Vehicle Concept for Fast Interplanetary Travel

    Science.gov (United States)

    Williams, Craig H.; Borowski, Stanley K.; Dudzinski, Leonard A.; Juhasz, Albert J.

    1998-01-01

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Initial requirements were for a human mission to Saturn with a greater than 5% payload mass fraction and a one way trip time of less than one year. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 235 days, with an initial mass in low Earth orbit of 2,941 mt. Engineering conceptual design, analysis, and assessment was performed on all ma or systems including payload, central truss, nuclear reactor (including divertor and fuel injector), power conversion (including turbine, compressor, alternator, radiator, recuperator, and conditioning), magnetic nozzle, neutral beam injector, tankage, start/re-start reactor and battery, refrigeration, communications, reaction control, and in-space operations. Detailed assessment was done on reactor operations, including plasma characteristics, power balance, power utilization, and component design.

  5. The ISS Water Processor Catalytic Reactor as a Post Processor for Advanced Water Reclamation Systems

    Science.gov (United States)

    Nalette, Tim; Snowdon, Doug; Pickering, Karen D.; Callahan, Michael

    2007-01-01

    Advanced water processors being developed for NASA s Exploration Initiative rely on phase change technologies and/or biological processes as the primary means of water reclamation. As a result of the phase change, volatile compounds will also be transported into the distillate product stream. The catalytic reactor assembly used in the International Space Station (ISS) water processor assembly, referred to as Volatile Removal Assembly (VRA), has demonstrated high efficiency oxidation of many of these volatile contaminants, such as low molecular weight alcohols and acetic acid, and is considered a viable post treatment system for all advanced water processors. To support this investigation, two ersatz solutions were defined to be used for further evaluation of the VRA. The first solution was developed as part of an internal research and development project at Hamilton Sundstrand (HS) and is based primarily on ISS experience related to the development of the VRA. The second ersatz solution was defined by NASA in support of a study contract to Hamilton Sundstrand to evaluate the VRA as a potential post processor for the Cascade Distillation system being developed by Honeywell. This second ersatz solution contains several low molecular weight alcohols, organic acids, and several inorganic species. A range of residence times, oxygen concentrations and operating temperatures have been studied with both ersatz solutions to provide addition performance capability of the VRA catalyst.

  6. Nonisothermal reactors for the production of pure water from peritoneal dialysis waste waters.

    Science.gov (United States)

    Diano, N; Ettari, G; Grano, V; Gaeta, F S; Rossi, S; Bencivenga, U; D'Alterio, C; Ruocco, G; Mita, L; De Santo, N G; Canciglia, P; Mita, D G

    2007-01-01

    The diffusion of peritoneal dialysis (PD) at home is somewhat restricted by the difficulty of transport and storage of a large amount of dialytic solutions. This problem is exacerbated in the case of hemodialysis. With the aim of producing pure water to be used in preparing the solution for peritoneal dialysis, or for hemodialysis in general, as one example, we purified the spent dialysate solution from PD. Experiments were carried out with 24 dialysate solutions taken from 8 patients. Pure water was obtained by means of a thermodialysis process in a hollow fiber reactor operating under nonisothermal conditions. Results show that the yield of the nonisothermal process is dependent on the temperature difference applied across the hydrophobic membranes. The production of pure water per square meter of membrane and per hour was equal to 0.55 or 1.2 or 2.0 liters, with a temperature difference of 11 degrees C or 21 degrees C or 28 degrees C, respectively. These results encourage the use of the thermodialysis process in the production of pure water for clinical uses.

  7. Co-digestion of concentrated black water and kitchen refuse in an accumulation system within the DESAR (decentralized sanitation and reuse) concept

    NARCIS (Netherlands)

    Kujawa, K.; Elmitwally, T.A.; Gaillard, A.; Leeuwen, van M.; Zeeman, G.

    2003-01-01

    Co-digestion of concentrated black water and kitchen refuse within the DESAR concept was the objective of this pilot research. The digestion took place in two, non-mixed accumulation reactors (AC1 and AC2) inoculated with digested primary sludge from a WWTP at a temperature of 20degreesC for a perio

  8. Water-responsive shape memory hybrid: Design concept and demonstration

    Directory of Open Access Journals (Sweden)

    2011-05-01

    Full Text Available Shape memory materials are featured by their ability to recover their original shapes when a particular stimulus, such as heat, light, magnetic field, even moisture/water, etc. is applied. However, it is not an easy task for non-professionals to synthesize a shape memory material which can meet all the requirements of a particular application. Even for professionals, like materials researchers, it could involve tedious trial and error procedures. In this paper, the concept of water-responsive shape memory hybrid is proposed and the advantages are demonstrated by two examples. The hybrid concept is versatile and can be easily accessed by those even without much polymer/chemistry background. Moreover, the performance of such hybrids can be well-predicted. This concept can be further extended into solvent-responsive shape memory hybrids, which can be routinely designed and realized in a Do-It-Yourself manner by almost anyone.

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

    Science.gov (United States)

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

    2011-07-01

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

  10. OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-07-01

    In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels.

  11. Analysis of Pressurized Water Reactor Primary Coolant Leak Events Caused by Thermal Fatigue

    Energy Technology Data Exchange (ETDEWEB)

    Atwood, Corwin Lee; Shah, Vikram Naginbhai; Galyean, William Jospeh

    1999-09-01

    We present statistical analyses of pressurized water reactor (PWR) primary coolant leak events caused by thermal fatigue, and discuss their safety significance. Our worldwide data contain 13 leak events (through-wall cracking) in 3509 reactor-years, all in stainless steel piping with diameter less than 25 cm. Several types of data analysis show that the frequency of leak events (events per reactor-year) is increasing with plant age, and the increase is statistically significant. When an exponential trend model is assumed, the leak frequency is estimated to double every 8 years of reactor age, although this result should not be extrapolated to plants much older than 25 years. Difficulties in arresting this increase include lack of quantitative understanding of the phenomena causing thermal fatigue, lack of understanding of crack growth, and difficulty in detecting existing cracks.

  12. Scrap tyre recycling process with molten zinc as direct heat transfer and solids separation fluid: A new reactor concept.

    Science.gov (United States)

    Riedewald, Frank; Goode, Kieran; Sexton, Aidan; Sousa-Gallagher, Maria J

    2016-01-01

    Every year about 1.5 billion tyres are discarded worldwide representing a large amount of solid waste, but also a largely untapped source of raw materials. The objective of the method was to prove the concept of a novel scrap tyre recycling process which uses molten zinc as the direct heat transfer fluid and, simultaneously, uses this media to separate the solids products (i.e. steel and rCB) in a sink-float separation at an operating temperature of 450-470 °C. This methodology involved: •construction of the laboratory scale batch reactor,•separation of floating rCB from the zinc,•recovery of the steel from the bottom of the reactor following pyrolysis.

  13. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  14. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  15. Component failures at pressurized water reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis.

  16. Design parameters for sludge reduction in an aquatic worm reactor

    NARCIS (Netherlands)

    Hendrickx, T.L.G.; Temmink, B.G.; Elissen, H.J.H.; Buisman, C.J.N.

    2010-01-01

    Reduction and compaction of biological waste sludge from waste water treatment plants (WWTPs) can be achieved with the aquatic worm Lumbriculus variegatus. In our reactor concept for a worm reactor, the worms are immobilised in a carrier material. The size of a worm reactor will therefore mainly be

  17. Space Molten Salt Reactor Concept for Nuclear Electric Propulsion and Surface Power

    Science.gov (United States)

    Eades, M.; Flanders, J.; McMurray, N.; Denning, R.; Sun, X.; Windl, W.; Blue, T.

    Students at The Ohio State University working under the NASA Steckler Grant sought to investigate how molten salt reactors with fissile material dissolved in a liquid fuel medium can be applied to space applications. Molten salt reactors of this kind, built for non-space applications, have demonstrated high power densities, high temperature operation without pressurization, high fuel burn up and other characteristics that are ideal for space fission systems. However, little research has been published on the application of molten salt reactor technology to space fission systems. This paper presents a conceptual design of the Space Molten Salt Reactor (SMSR), which utilizes molten salt reactor technology for Nuclear Electric Propulsion (NEP) and surface power at the 100 kWe to 15 MWe level. Central to the SMSR design is a liquid mixture of LiF, BeF2 and highly enriched U235F4 that acts as both fuel and core coolant. In brief, some of the positive characteristics of the SMSR are compact size, simplified core design, high fuel burn up percentages, proliferation resistant features, passive safety mechanisms, a considerable body of previous research, and the possibility for flexible mission architecture.

  18. Coolability of corium debris under severe accident conditions in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Saidur

    2013-11-15

    The debris bed which may be formed in different stages of a severe accident will be hot and heated by decay heat from the radioactive fission products. In order to establish a steady state of long-term cooling, this hot debris needs to be quenched at first. If quenching by water ingression into the dry bed is not rapid enough then heat-up by decay heat in still dry regions may again yield melting. Thus, chances of coolability must be investigated considering quenching against heat-up due to decay heat, in the context of reactor safety research. As a basis of the present investigations, models for simulation of two phase flow through porous medium were already available in the MEWA code, being under development at IKE. The objective of this thesis is to apply the code in essential phases of severe accidents and to investigate the chances, options and measures for coolability. Further, within the tasks, improvements to remove weaknesses in modeling and implementation of extensions concerning missing parts are included. It was identified previously that classical models without explicit considering the interfacial friction, can predict dryout heat flux (DHF) well under top fed condition but under-predict DHF values under bottom flooding conditions. Tung and Dhir introduced an interfacial friction term in their model, but this model has deficits for smaller particles considered as relevant for reactor conditions. Therefore, some modification of Tung and Dhir model is proposed in the present work to extent it for smaller particles. A significant improvement with the new friction description (Modified Tung and Dhir, MTD) is obtained considering the aim of a unified description for both top and bottom flooding conditions and for broad bandwidth of bed conditions. Calculations for reactor conditions are carried out in order to explore whether or to which degree coolability can be concluded, how strong the trend to coolability is and where major limits occur. The general

  19. On the study of catalytic membrane reactor for water detritiation: Modeling approach

    Energy Technology Data Exchange (ETDEWEB)

    Liger, Karine, E-mail: karine.liger@cea.fr [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Mascarade, Jérémy [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France); Joulia, Xavier; Meyer, Xuan-Mi [Université de Toulouse, INPT, UPS, Laboratoire de Génie Chimique, 4, Allée Emile Monso, Toulouse F-31030 (France); CNRS, Laboratoire de Génie Chimique, Toulouse F-31030 (France); Troulay, Michèle; Perrais, Christophe [CEA, DEN, DTN/SMTA/LIPC Cadarache, Saint Paul-lez-Durance F-13108 (France)

    2016-11-01

    Highlights: • Experimental results for the conversion of tritiated water (using deuterium as a simulant of tritium) by means of a catalytic membrane reactor in view of tritium recovery. • Phenomenological 2D model to represent catalytic membrane reactor behavior including the determination of the compositions of gaseous effluents. • Good agreement between the simulation results and experimental measurements performed on the dedicated facility. • Explanation of the unexpected behavior of the catalytic membrane reactor by the modeling results and in particular the gas composition estimation. - Abstract: In the framework of tritium recovery from tritiated water, efficiency of packed bed membrane reactors have been successfully demonstrated. Thanks to protium isotope swamping, tritium bonded water can be recovered under the valuable Q{sub 2} form (Q = H, D or T) by means of isotope exchange reactions occurring on catalyst surface. The use of permselective Pd-based membrane allows withdrawal of reactions products all along the reactor, and thus limits reverse reaction rate to the benefit of the direct one (shift effect). The reactions kinetics, which are still little known or unknown, are generally assumed to be largely greater than the permeation ones so that thermodynamic equilibriums of isotope exchange reactions are generally assumed. This paper proposes a new phenomenological 2D model to represent catalytic membrane reactor behavior with the determination of gas effluents compositions. A good agreement was obtained between the simulation results and experimental measurements performed on a dedicated facility. Furthermore, the gas composition estimation permits to interpret unexpected behavior of the catalytic membrane reactor. In the next future, further sensitivity analysis will be performed to determine the limits of the model and a kinetics study will be conducted to assess the thermodynamic equilibrium of reactions.

  20. Construction of the advanced boiling water reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Natsume, Nobuo; Noda, Hiroshi [Tokyo Electric Power Co. (Japan). Nuclear Power Plant Construction Dept.

    1996-07-01

    The Advanced Boiling Reactor (ABWR) has been developed with international cooperation between Japan and the US as the generation of plants for the 1990s and beyond. It incorporates the best BWR technologies from the world in challengeable pursuit of improved safety and reliability, reduced construction and operating cost, reduced radiation exposure and radioactive waste. Tokyo Electric Power Company (MPCO) decided to apply the first ABWRs to unit No. 6 and 7 of Kashiwazaki-Kariwa nuclear power station (K-6 and 7). These units are scheduled to commence commercial operation in December 1996 and July 1997 respectively. Particular attention is given in this discussion to the construction period from rock inspection for the reactor building to commercial operation, which is to be achieved in only 52 months through innovative and challenging construction methods. To date, construction work is advancing ahead of the original schedule. This paper describes not only how to shorten the construction period by adoption of a variety of new technologies, such as all-weather construction method and large block module construction method, but also how to check and test the state of the art technologies during manufacturing and installation of new equipment for K-6 and 7.

  1. Local stability tests in Dresden 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    March-Leuba, J.; Fry, D.N.; Buchanan, M.E.; McNew, C.O.

    1984-04-01

    This report presents the results of a local stability test performed at Dresden Unit 2 in May 1983 to determine the effect of a new fuel element design on local channel stability. This test was performed because the diameter of the new fuel rods increases the heat transfer coefficient, making the reactor more responsive and, thus, more susceptible to instabilities. After four of the new fuel elements with a 9 x 9 array of fuel rods were loaded into Dresden 2, the test was performed by inserting an adjacent control rod all the way in and then withdrawing it to its original position at maximum speed. At the moment of the test, reactor conditions were 52.7% power and 38.9% flow. Both the new 9 x 9 fuel elements and the standard 8 x 8 ones proved to be locally stable when operating at minimum pump speed at the beginning of cycle in Dresden 2, and no significant difference was found between the behavior of the two fuel types. Finally, Dresden 2 showed a high degree of stability during control rod and normal noise type perturbations.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-06

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-06

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

  4. Reconstructing the direction of reactor antineutrinos via electron scattering in Gd-doped water Cherenkov detectors

    Science.gov (United States)

    Hellfeld, D.; Bernstein, A.; Dazeley, S.; Marianno, C.

    2017-01-01

    The potential of elastic antineutrino-electron scattering in a Gd-doped water Cherenkov detector to determine the direction of a nuclear reactor antineutrino flux was investigated using the recently proposed WATCHMAN antineutrino experiment as a baseline model. The expected scattering rate was determined assuming a 13-km standoff from a 3.758-GWt light water nuclear reactor and the detector response was modeled using a Geant4-based simulation package. Background was estimated via independent simulations and by scaling published measurements from similar detectors. Background contributions were estimated for solar neutrinos, misidentified reactor-based inverse beta decay interactions, cosmogenic radionuclides, water-borne radon, and gamma rays from the photomultiplier tubes (PMTs), detector walls, and surrounding rock. We show that with the use of low background PMTs and sufficient fiducialization, water-borne radon and cosmogenic radionuclides pose the largest threats to sensitivity. Directional sensitivity was then analyzed as a function of radon contamination, detector depth, and detector size. The results provide a list of experimental conditions that, if satisfied in practice, would enable antineutrino directional reconstruction at 3σ significance in large Gd-doped water Cherenkov detectors with greater than 10-km standoff from a nuclear reactor.

  5. Environmentally assisted cracking in light water reactors. Semiannual progress report, January 1996--June 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E. [and others

    1997-05-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from January 1996 to June 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, and (c) EAC of Alloys 600 and 690. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated boiling water reactor (BWR) water at 288{degrees}C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in air and high-purity, low-DO water. 83 refs., 60 figs., 14 tabs.

  6. Environmentally assisted cracking in light water reactors. Semiannual report, April 1994--September 1994, Volume 19

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [and others

    1995-09-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors from April to September 1994. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in piping and reactor pressure vessels, (b) EAC of austenitic stainless steels (SSs) and Alloy 600, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests have been conducted on A106-Gr B and A533-Gr B steels in oxygenated water to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack growth data were obtained on fracture-mechanics specimens of SSs and Alloy 600 to investigate EAC in simulated boiling water reactor (BWR) and pressurized water reactor environments at 289{degrees}C. The data were compared with predictions from crack growth correlations developed at ANL for SSs in water and from rates in air from Section XI of the ASME Code. Microchemical changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  7. Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1994-06-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289{degree}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  8. Environmentally assisted cracking in light water reactors. Semiannual report July 1996--December 1996

    Energy Technology Data Exchange (ETDEWEB)

    Chopra, O.K.; Chung, H.M.; Gavenda, D.J. [Argonne National Lab., IL (United States)] [and others

    1997-10-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors from July 1996 to December 1996. Topics that have been investigated include (a) fatigue of carbon, low-alloy, and austenitic stainless steels (SSs) used in reactor piping and pressure vessels, (b) irradiation-assisted stress corrosion cracking of Type 304 SS, (c) EAC of Alloy 600, and (d) characterization of residual stresses in welds of boiling water reactor (BWR) core shrouds by numerical models. Fatigue tests were conducted on ferritic and austenitic SSs in water that contained various concentrations of dissolved oxygen to determine whether a slow strain rate applied during various portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Slow-strain-rate-tensile tests were conducted in simulated BWR water at 288 C on SS specimens irradiated to a low fluence in the Halden reactor and the results were compared with similar data from a control-blade sheath and neutron-absorber tubes irradiated in BWRs to the same fluence level. Crack-growth-rate tests were conducted on compact-tension specimens from a low-carbon content heat of Alloy 600 in high-purity oxygenated water at 289 C. Residual stresses and stress intensity factors were calculated for BWR core shroud welds.

  9. Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O2 cannot be ignored, especially for the FHR, in which environment the product, SiO2, can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

  10. Capabilities of the ATHENA computer code for modeling the SP-100 space reactor concept

    Science.gov (United States)

    Fletcher, C. D.

    1985-09-01

    The capability to perform thermal-hydraulic analyses of an SP-100 space reactor was demonstrated using the ATHENA computer code. The preliminary General Electric SP-100 design was modeled using Athena. The model simulates the fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of this design. Two ATHENA demonstration calculations were performed simulating accident scenarios. A mask for the SP-100 model and an interface with the Nuclear Plant Analyzer (NPA) were developed, allowing a graphic display of the calculated results on the NPA.

  11. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gavilian-Moreno, Carlos [Iberdrola Generacion, S.A., Cofrentes Nuclear Power Plant, Project Engineering Department, Paraje le Plano S/N, Valencia (Spain); Espinosa-Paredes, Gilberto [Area de ingeniera en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Mexico city (Mexico)

    2016-04-15

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  12. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  13. Nanostructure of Metallic Particles in Light Water Reactor Used Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mausolf, Edward J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mcnamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Schwantes, Jon M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-11

    The extraordinary nano-structure of metallic particles in light water reactor fuels points to possible high reactivity through increased surface area and a high concentration of high energy defect sites. We have analyzed the metallic epsilon particles from a high burn-up fuel from a boiling water reactor using transmission electron microscopy and have observed a much finer nanostructure in these particles than has been reported previously. The individual round particles that varying in size between ~20 and ~50 nm appear to consist of individual crystallites on the order of 2-3 nm in diameter. It is likely that in-reactor irradiation induce displacement cascades results in the formation of the nano-structure. The composition of these metallic phases is variable yet the structure of the material is consistent with the hexagonal close packed structure of epsilon-ruthenium. These findings suggest that unusual catalytic behavior of these materials might be expected, particularly under accident conditions.

  14. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  15. Porous Photocatalytic Membrane Microreactor (P2M2): A new reactor concept for photochemistry

    NARCIS (Netherlands)

    Aran, H.C.; Salamon, D.; Rijnaarts, T.; Mul, G.; Wessling, M.; Lammertink, R.G.H.

    2011-01-01

    In this study, a new membrane microreactor concept for multiphase photocatalytic reactions is demonstrated. Microfabrication, photocatalyst immobilization and surface modification steps were performed to develop a Porous Photocatalytic Membrane Microreactor (P2M2). This concept benefits from a stabl

  16. Porous Photocatalytic Membrane Microreactor (P2M2): A new reactor concept for photochemistry

    NARCIS (Netherlands)

    Aran, H.C.; Salamon, David; Rijnaarts, Timon; Rijnaarts, T.; Mul, Guido; Wessling, Matthias; Lammertink, Rob G.H.

    2011-01-01

    In this study, a new membrane microreactor concept for multiphase photocatalytic reactions is demonstrated. Microfabrication, photocatalyst immobilization and surface modification steps were performed to develop a Porous Photocatalytic Membrane Microreactor (P2M2). This concept benefits from a

  17. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  18. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... 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 considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  19. Hydraulic performance of a proposed in situ photocatalytic reactor for degradation of MTBE in water.

    Science.gov (United States)

    Lim, Leonard Lik Pueh; Lynch, Rod

    2011-01-01

    Methyl tert-butyl ether (MTBE) groundwater remediation projects often require a combination of technologies resulting in increasing the project costs. A cost-effective in situ photocatalytic reactor design, Honeycomb II, is proposed and tested for its efficiency in MTBE degradation at various flows. This study is an intermediate phase of the research in developing an in situ photocatalytic reactor for groundwater remediation. It examines the effect of the operating variables: air and water flow and double passages through Honeycomb II, on the MTBE removal. MTBE vaporisation is affected by not only temperature, Henry's law constant and air flow to volume ratio but also reactor geometry. The column reactor achieved more than 84% MTBE removal after 8 h at flows equivalent to horizontal groundwater velocities slower than 21.2 cm d⁻¹. Despite the contrasting properties between a photocatalytic indicator methylene blue and MTBE, the reactor efficiency in degrading both compounds showed similar responses towards flow (equivalent groundwater velocity and hydraulic residence time (HRT)). The critical HRT for both compounds was approximately 1 d, which corresponded to a velocity of 21.2 cm d⁻¹. A double pass through both new and used catalysts achieved more than 95% MTBE removal after two passes in 48 h. It also verified that the removal efficiency can be estimated via the sequential order of the removal efficiency of one pass obtained in the laboratory. This study reinforces the potential of this reactor design for in situ groundwater remediation.

  20. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Popp, M

    1987-01-01

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated.

  1. Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, M.A.; Werner, R.W.; Carlson, G.A.; Cornish, D.N.

    1976-04-01

    Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements. Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated.

  2. Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1995-03-01

    This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials.

  3. TMI-2 Reactor Building source term measurements: surfaces and basement water and sediment

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C V; Keefer, D G

    1984-10-01

    Presented in this report are the results of radiochemical and elemental analyses performed on samples collected from the Three Mile Island Unit 2 Reactor Building from August 1979 to December 1983. The quantities of fission products and core materials that were measured on the external surfaces in the Reactor Building or in the water and sediment in its basement are summarized. Recent analysis results for access panels removed from the air cooling assembly and for liquid and particulate samples collected from the Reactor Building sump and reactor coolant drain tank are included in the report. Measurements show that 59% of the /sup 3/H, 2.7% of the /sup 90/Sr, 15% of the /sup 129/I, 20% of the /sup 131/I, and 42% of the /sup 137/Cs originally in the core at the time of the accident could be accounted for outside the core in the Reactor Building. With the exceptions of /sup 90/Sr and /sup 144/Ce, the vast majority of each radionuclide released was found dispersed in the water and sediment in the basement.

  4. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  5. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

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

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

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

  7. 77 FR 16270 - Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water...

    Science.gov (United States)

    2012-03-20

    ... COMMISSION Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water... license renewal interim staff guidance (LR-ISG), LR-ISG-2011-04, ``Updated Aging Management Criteria for... Aging Lessons Learned (GALL) Report for the aging management of stainless steel structures...

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

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

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

  9. Neutron fluence depth profiles in water phantom on epithermal beam of LVR-15 research reactor.

    Science.gov (United States)

    Viererbl, L; Klupak, V; Lahodova, Z; Marek, M; Burian, J

    2010-01-01

    Horizontal channel with epithermal neutron beam at the LVR-15 research reactor is used mainly for boron neutron capture therapy. Neutron fluence depth profiles in a water phantom characterise beam properties. The neutron fluence (approximated by reaction rates) depth profiles were measured with six different types of activation detectors. The profiles were determined for thermal, epithermal and fast neutrons.

  10. Water plant modifications for increased production at B, C, D, DR, F, and H Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, L.B.; Corley, J.P.

    1960-04-15

    The purpose of this report is to define the extent of modifications necessary to increase capacities of the 100-B, C, D, DR, F, and H water plants for reactor flows of 90,000 95,000 105,000 and 115,000 GPM, and to provide supporting data for budget studies for increased production.

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

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

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

  12. 77 FR 38339 - Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security...

    Science.gov (United States)

    2012-06-27

    ... COMMISSION Dairyland Power Cooperative, La Crosse Boiling Water Reactor Exemption From Certain Security... Dairyland Power Cooperative (DPC). The LACBWR was a nuclear power plant of nominal 50 Mw electrical output... from the regulations in part 73 as it determines are authorized by law and will not endanger life...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  14. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR.

  15. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-1: Pressurized Water Reactors.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical pressurized water reactor (PWR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module is the PWR…

  16. Design characteristics for pressurized water small modular nuclear power reactors with focus on safety

    Energy Technology Data Exchange (ETDEWEB)

    Kani, Iraj Mahmoudzadeh [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; Zandieh, Mehdi [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty; Abadi, Saeed Kheirollahi Hossein [International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty

    2016-05-15

    Small Modular Reactors (SMRs) are a technology, attracting attention. Light water SMR possess an upgraded design case and emphasize the significance of integral models. Beside of these advantages, SMRs has faced numerous challenges, e.g. licensing, cost/investment, safety and security observation, social and environmental issues in building new plants.

  17. Environmentally assisted cracking in light water reactors. Semiannual report, April--September 1991: Volume 13

    Energy Technology Data Exchange (ETDEWEB)

    Kassner, T F; Ruther, W E; Chung, H M; Hicks, P D; Hins, A G; Park, J Y; Soppet, W K; Shack, W J [Argonne National Lab., IL (United States)

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with {approx} 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289{degrees}C.

  18. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soon Ho; Kim, Dae Seop; Kim, Jae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2014-06-15

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

  19. Supercritical water oxidation - Concept analysis for evolutionary Space Station application

    Science.gov (United States)

    Hall, John B., Jr.; Brewer, Dana A.

    1986-01-01

    The ability of a supercritical water oxidation (SCWO) concept to reduce the number of processes needed in an evolutionary Space Station design's Environmental Control and Life Support System (ECLSS), while reducing resupply requirements and enhancing the integration of separate ECLSS functions into a single Supercritical Water Oxidation process, is evaluated. While not feasible for an initial operational capability Space Station, the SCWO's application to the evolutionary Space Station configuration would aid the integration of eight ECLSS functions into a single one, thereby significantly reducing program costs.

  20. Mitigation of Oil in Water Column: Concept Development

    Science.gov (United States)

    2016-06-01

    Project No. 4702 7. Author(s) Alexander Balsley, Dr. Michele Fitzpatrick, Peter A. Tebeau 8. Performing Report No. RDC UDI # 1291 9...of Mexico . Figure A-4. Sorbent pom-poms from the Detroit River 1996. Figure A-5. Vessel-Submerged Oil Recovery System. Mitigation of Oil in... Mexico Droplets from chemical and possibly natural dispersion. None Mitigation of Oil in Water Column: Concept Development B-2 UNCLAS//Public

  1. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

  2. The specifications of the European electricians for the conception of the future reactors; Le cahier des charges des electriciens europeens pour la conception des futurs reacteurs (les EUR)

    Energy Technology Data Exchange (ETDEWEB)

    Berbey, P. [Electricite de France, 92 - Clamart (France). Service Etudes et Projets Thermiques et Nucleaires

    2007-09-15

    The cooperation between European electricians through European utility requirements (E.U.R.) continues since the beginning of the 90. Transplanted initially on the program A.L.W.R., the E.U.R. project gradually managed to bring to the foreground a common specification for light water type reactors of 3. generation to be built in Europe. The nuclear safety was one of the major axes of development and the set of requirements of E.U.R. safety establishes one of European standard today. A continuous dialogue with the other actors of the nuclear power at the world level allowed to consolidate the proposed options. (N.C.)

  3. Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor.

    Science.gov (United States)

    Polo-López, M I; Fernández-Ibáñez, P; Ubomba-Jaswa, E; Navntoft, C; García-Fernández, I; Dunlop, P S M; Schmid, M; Byrne, J A; McGuigan, K G

    2011-11-30

    Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

  4. Elimination of water pathogens with solar radiation using an automated sequential batch CPC reactor

    Energy Technology Data Exchange (ETDEWEB)

    Polo-Lopez, M.I., E-mail: mpolo@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Fernandez-Ibanez, P., E-mail: pilar.fernandez@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Ubomba-Jaswa, E., E-mail: euniceubombajaswa@yahoo.com [Natural Resources and the Environment, CSIR, PO Box 395, Pretoria (South Africa); Navntoft, C., E-mail: christian.navntoft@solarmate.com.ar [Instituto de Investigacion e Ingenieria Ambiental, Universidad Nacional de San Martin (3iA-UNSAM), Peatonal Belgrano 3563, B1650ANQ San Martin (Argentina); Universidad Tecnologica Nacional - Facultad Regional Buenos Aires - Departamento de Ingenieria Civil - Laboratorio de Estudios sobre Energia Solar, (UTN-FRBA-LESES), Mozart 2300, (1407) Ciudad Autonoma de Buenos Aires, Republica Argentina (Argentina); Garcia-Fernandez, I., E-mail: irene.garcia@psa.es [Plataforma Solar de Almeria - CIEMAT, PO Box 22, 04200 Tabernas, Almeria (Spain); Dunlop, P.S.M., E-mail: psm.dunlop@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Schmid, M. [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); Byrne, J.A., E-mail: j.byrne@ulster.ac.uk [Department of Physiology and Medical Physics, Royal College of Surgeons in Ireland, Dublin 2 (Ireland); and others

    2011-11-30

    Solar disinfection (SODIS) of water is a well-known, effective treatment process which is practiced at household level in many developing countries. However, this process is limited by the small volume treated and there is no indication of treatment efficacy for the user. Low cost glass tube reactors, together with compound parabolic collector (CPC) technology, have been shown to significantly increase the efficiency of solar disinfection. However, these reactors still require user input to control each batch SODIS process and there is no feedback that the process is complete. Automatic operation of the batch SODIS process, controlled by UVA-radiation sensors, can provide information on the status of the process, can ensure the required UVA dose to achieve complete disinfection is received and reduces user work-load through automatic sequential batch processing. In this work, an enhanced CPC photo-reactor with a concentration factor of 1.89 was developed. The apparatus was automated to achieve exposure to a pre-determined UVA dose. Treated water was automatically dispensed into a reservoir tank. The reactor was tested using Escherichia coli as a model pathogen in natural well water. A 6-log inactivation of E. coli was achieved following exposure to the minimum uninterrupted lethal UVA dose. The enhanced reactor decreased the exposure time required to achieve the lethal UVA dose, in comparison to a CPC system with a concentration factor of 1.0. Doubling the lethal UVA dose prevented the need for a period of post-exposure dark inactivation and reduced the overall treatment time. Using this reactor, SODIS can be automatically carried out at an affordable cost, with reduced exposure time and minimal user input.

  5. Non normal modal analysis of oscillations in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Suarez-Antola, Roberto, E-mail: roberto.suarez@miem.gub.uy [Ministerio de Industria, Energia y Mineria (MIEM), Montevideo (Uruguay); Flores-Godoy, Jose-Job, E-mail: job.flores@ibero.mx [Universidad Iberoamericana (UIA), Mexico, DF (Mexico). Dept. de Fisica Y Matematicas

    2013-07-01

    The first objective of the present work is to construct a simple reduced order model for BWR stability analysis, combining a two nodes nodal model of the thermal hydraulics with a two modes modal model of the neutronics. Two coupled non-linear integral-differential equations are obtained, in terms of one global (in phase) and one local (out of phase) power amplitude, with direct and cross feedback reactivities given as functions of thermal hydraulics core variables (void fractions and temperatures). The second objective is to apply the effective life time approximation to further simplify the nonlinear equations. Linear approximations for the equations of the amplitudes of the global and regional modes are derived. The linearized equation for the amplitude of the global mode corresponds to a decoupled and damped harmonic oscillator. An analytical closed form formula for the damping coefficient, as a function of the parameters space of the BWR, is obtained. The coefficient changes its sign (with the corresponding modification in the decay ratio) when a stability boundary is crossed. This produces a supercritical Hopf bifurcation, with the steady state power of the reactor as the bifurcation parameter. However, the linearized equation for the amplitude of the regional mode corresponds always to an over-damped and always coupled (with the amplitude of the global mode) harmonic oscillator, for every set of possible values of core parameters (including the steady state power of the reactor) in the framework of the present mathematical model. The equation for the above mentioned over damped linear oscillator is closely connected with a non-normal operator. Due to this connection, there could be a significant transient growth of some solutions of the linear equation. This behavior allows a significant shrinking of the basin of attraction of the equilibrium state. The third objective is to apply the above approach to partially study the stability of the regional mode and

  6. Experimental techniques to determine salt formation and deposition in supercritical water oxidation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chan, J.P.C.; LaJeunesse, C.A.; Rice, S.F.

    1994-08-01

    Supercritical Water Oxidation (SCWO) is an emerging technology for destroying aqueous organic waste. Feed material, containing organic waste at concentrations typically less than 10 wt % in water, is pressurized and heated to conditions above water`s critical point where the ability of water to dissolve hydrocarbons and other organic chemicals is greatly enhanced. An oxidizer, is then added to the feed. Given adequate residence time and reaction temperature, the SCWO process rapidly produces innocuous combustion products. Organic carbon and nitrogen in the feed emerge as CO{sub 2} and N{sub 2}; metals, heteroatoms, and halides appear in the effluent as inorganic salts and acids. The oxidation of organic material containing heteroatoms, such as sulfur or phosphorous, forms acid anions. In the presence of metal ions, salts are formed and precipitate out of the supercritical fluid. In a tubular configured reactor, these salts agglomerate, adhere to the reactor wall, and eventually interfere by causing a flow restriction in the reactor leading to an increase in pressure. This rapid precipitation is due to an extreme drop in salt solubility that occurs as the feed stream becomes supercritical. To design a system that can accommodate the formation of these salts, it is important to understand the deposition process quantitatively. A phenomenological model is developed in this paper to predict the time that reactor pressure begins to rise as a function of the fluid axial temperature profile and effective solubility curve. The experimental techniques used to generate effective solubility curves for one salt of interest, Na{sub 2}SO{sub 4}, are described, and data is generated for comparison. Good correlation between the model and experiment is shown. An operational technique is also discussed that allows the deposited salt to be redissolved in a single phase and removed from the affected portion of the reactor. This technique is demonstrated experimentally.

  7. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  8. The controllability analysis of the purification system for heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. D.; Cho, B. H.; Shin, C. H.; Kim, S. H. [KEPRI, Taejon (Korea, Republic of); Lee, Y. K.; Kim, K. U. [KHNP, Kyungju (Korea, Republic of)

    2001-10-01

    The heavy water reactor such as Wolsung No.1 and No.2 has a purification system to purify the reactor coolant. The control system regulates the coolant temperature to protect the ion exchanger. After the fuel exchanges of operating plant, the increase of the coolant pressure makes the purification temperature control difficult. In this paper, the controllability of the control dynamics of the purification system was analysed and the optimal parameters were proposed. To reduce the effects of the flow disturbance, the feedforward control structure was proposed and analysed.

  9. A review of qualitative inspection aspects of end fittings in an Indian pressurized heavy water reactor

    Directory of Open Access Journals (Sweden)

    Urva Pancholi

    2016-07-01

    Full Text Available The paper provides a summarized description of the current state of knowledge and practices used in India, in the qualitative inspection of end fittings – a key component of the fuel channel assembly of a pressurized heavy water reactor (PHWR, generally of a Canadian Deuterium Uranium (CANDU type. Further it discusses various quality inspection techniques; and the high standards and mechanical precision of the job required, to be accepted as viable nuclear reactor component. The techniques, instruments and specific data for such components mentioned here are synthesized results from primary research and knowledge available in this area, in order to produce coherent argument focused on quality control of end fittings.

  10. Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

    Energy Technology Data Exchange (ETDEWEB)

    Rice, R.E.

    1976-09-01

    Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments.

  11. In-Reactor Oxidation of Zircaloy-4 Under Low Water Vapor Pressures

    Energy Technology Data Exchange (ETDEWEB)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin; Longhurst, Glen

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330° and 370°C). Data from these tests will be used to support fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex- reactor test results were performed to evaluate the influence of irradiation.

  12. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    Energy Technology Data Exchange (ETDEWEB)

    Luscher, Walter G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Senor, David J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clayton, Kevin K. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Longhurst, Glen R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ºC). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  13. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W [Westinghouse Savannah River Co., Aiken, SC (USA); Hagrman, D L [EG and G Idaho, Inc., Idaho Falls, ID (USA); McClure, P R; Leonard, M T [Science Applications International Corp., Albuquerque, NM (USA)

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  14. Degraded core analysis for the pressurized-water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-02-09

    An analysis of the likelihood and the consequences of 'degraded-core accidents' has been undertaken for the proposed Sizewell B PWR. In such accidents, degradation of the core geometry occurs as a result of overheating. Radionuclides are released and may enter the environment, causing harmful effects. The analysis concludes that degraded-core accidents are highly improbable, the plant having been designed to reduce the frequency of such accidents to a value of order 10/sup -6/ per year. Tbe building containing the reactor would only fail in a small proportion of degraded-core accidents. In the great majority of cases the containment would remain intact and the release of radioactivity to the environment would be small. The risk to individuals have been calculated for both immediate and long term effects. Although the estimates of risk are approximate, studies to investigate the uncertainties, and sensitivities to different assumptions, show that potential errors are small compared with the very large 'margin of safety' between the risks estimated for Sizewell B and those that already exist in society.

  15. Biological mine water treatment operating a one stage reactor system

    CSIR Research Space (South Africa)

    Baloyi, MJ

    2006-05-01

    Full Text Available Mine drainage arises from oxidation of pyrites, due to exposure to air and water. Acid mine drainage normally contains high concentrations of sulphate, metals and acidity. These pollutants can be reduced by applying the biological sulphate reduction...

  16. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-10-03

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in ancircular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mas velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel.

  17. Overview of the US Department of Energy Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    K. A. McCarthy; D. L. Williams; R. Reister

    2012-05-01

    The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

  18. The pluralistic water research concept - a new human-water system research approach

    Science.gov (United States)

    Evers, Mariele; Höllermann, Britta; Almoradie, Adrian; Taft, Linda; Garcia-Santos, Glenda

    2017-04-01

    Sustainable water resources management has been and still is a main challenge for decision makers even though for the past number of decades integrative approaches and concepts (e.g. Integrated Water Resources Management - IWRM) have been developed to address problems on floods, droughts, water quality, water quantity, environment and ecology. Although somehow these approaches are aiming to address water related problems in an integrative approach and to some extent include or involve society in the planning and management, they still lack some of the vital components in including the social dimensions and their interaction with water. Understanding these dynamics in a holistic way and how they are shaped by time and space may tackle these shortcomings and provide more effective and sustainable management solutions with respect to a set of potential present social actions and values as well as possible futures. This paper aims to discuss challenges to coherently and comprehensively integrate the social dimensions of different human-water concepts like IWRM, socio-hydrology and waterscape. Against this background it will develop criteria for an integrative approach and present a newly developed concept termed pluralistic water research (PWR) concept. PWR is not only a pluralistic but also an integrative and interdisciplinary approach to acknowledge the social and water dimensions and their interaction and dynamics by considering more than one perspective of a water-related issue, hereby providing a set of multiple (future) developments. Our PWR concept will be illustrated by a case study application of the Canary island La Gomera. Furthermore an outlook on further possible developments of the PWR concept will be presented and discussed.

  19. Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept

    Science.gov (United States)

    Wenzler, C. J.

    1972-01-01

    A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.

  20. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    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

  1. Concept of a BNCT line with in-pool fission converter at MARIA reactor in Swierk

    Science.gov (United States)

    Pytel, Krzysztof; Andrzejewski, Krzysztof; Golnik, Natalia; Osko, Jakub

    2009-01-01

    BNCT facility in the Institute of Atomic Energy in Otwock-Swierk is under construction at the horizontal channel H2 of the research reactor MARIA. Measurements of the neutron energy spectrum performed at the front of the H2 experimental channel, have shown that flux of epithermal neutrons (above 10 keV) at the BNCT irradiation port was below 109 n cm-2 s-1 i.e. it was too low to be directly used for the BNCT treatment. Therefore, a fission converter will be placed between the reactor core and the periphery of the graphite reflector of MARIA reactor. The uranium converter will be powered by the densely packed EK-10 fuel elements with 10% enrichment. Preliminary calculations have shown that the total neutron flux in the converter will be about 1013 n cm-2 s-1 and flux of epithermal neutrons at the entrance to the filter/moderator of the beam will be about 2·1013 n cm-2 s-1.

  2. Materials for DEMO and reactor applications—boundary conditions and new concepts

    Science.gov (United States)

    Coenen, J. W.; Antusch, S.; Aumann, M.; Biel, W.; Du, J.; Engels, J.; Heuer, S.; Houben, A.; Hoeschen, T.; Jasper, B.; Koch, F.; Linke, J.; Litnovsky, A.; Mao, Y.; Neu, R.; Pintsuk, G.; Riesch, J.; Rasinski, M.; Reiser, J.; Rieth, M.; Terra, A.; Unterberg, B.; Weber, Th; Wegener, T.; You, J.-H.; Linsmeier, Ch

    2016-02-01

    DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER towards realizing fusion. For the realization of fusion energy especially, materials questions pose a significant challenge already today. Heat, particle and neutron loads are a significant problem to material lifetime when extrapolating to DEMO. For many of the issues faced, advanced materials solutions are under discussion or already under development. In particular, components such as the first wall and the divertor of the reactor can benefit from introducing new approaches such as composites or new alloys into the discussion. Cracking, oxidation as well as fuel management are driving issues when deciding for new materials. Here {{{W}}}{{f}}/{{W}} composites as well as strengthened CuCrZr components together with oxidation resilient tungsten alloys allow the step towards a fusion reactor. In addition, neutron induced effects such as transmutation, embrittlement and after-heat and activation are essential. Therefore, when designing a component an approach taking into account all aspects is required.

  3. A new concept of high flow rate non-thermal plasma reactor for air treatment

    Energy Technology Data Exchange (ETDEWEB)

    Goujard, V.; Tatibouet, J.M. [Univ. de Poitiers, Poitiers (France). Centre national de la recherche scientifique, Laboratoire de Catalyse en Chimie Organique

    2010-07-01

    Although several non-thermal plasma reactors have been tested for air treatment at the laboratory scale, up-scaling to pilot or industrial scale remains a challenge because several parameters must be considered, such as hydrodynamic behaviour, maximum voltage in an industrial environment, and maintenance of the system. This paper presented a newly developed reactor which consists to a DBD plasma generated on individual supports that could be directly inserted in gas pipes where air flow must be treated. Elimination of 40 percent of 15 ppm of propene was obtained with a energy density as low as 10 J/L. The propene conversion increased when a manganese oxide based catalyst was used because the ozone produced by the plasma was used as an as an oxidant. A simple model of the plasma-catalyst reactor behaviour showed that more than 90 percent of propene conversion can be expected for an input energy density of 10 J/L and residual ozone concentration less than 100 ppb.

  4. Quantifying the human impact on water resources: a critical review of the water footprint concept

    Science.gov (United States)

    Chenoweth, J.; Hadjikakou, M.; Zoumides, C.

    2014-06-01

    The water footprint is a consumption-based indicator of water use, referring to the total volume of freshwater used directly and indirectly by a nation or a company, or in the provision of a product or service. Despite widespread enthusiasm for the development and use of water footprints, some concerns have been raised about the concept and its usefulness. A variety of methodologies have been developed for water footprinting which differ with respect to how they deal with different forms of water use. The result is water footprint estimates which vary dramatically, often creating confusion. Despite these methodological qualms, the concept has had notable success in raising awareness about water use in agricultural and industrial supply chains, by providing a previously unavailable and (seemingly) simple numerical indicator of water use. Nevertheless, and even though a range of uses have already been suggested for water footprinting, its policy value remains unclear. Unlike the carbon footprint which provides a universal measure of human impact on the atmosphere's limited absorptive capacity, the water footprint in its conventional form solely quantifies a single production input without any accounting of the impacts of use, which vary spatially and temporally. Following an extensive review of the literature related to water footprints, this paper critically examines the present uses of the concept, focusing on its current strengths, shortcomings and promising research avenues to advance it.

  5. Preliminary Feasibility Testing of the BRIC Brine Water Recovery Concept

    Science.gov (United States)

    Callahan, Michael R.; Pensinger, Stuart; Pickering, Karen D.

    2011-01-01

    The Brine Residual In-Containment (BRIC) concept was developed as a new technology to recover water from spacecraft wastewater brines. Such capability is considered critical to closing the water loop and achieving a sustained human presence in space. The intention of the BRIC concept is to increase the robustness and efficiency of the dewatering process by performing drying inside the container used for the final disposal of the residual brine solid. Recent efforts in the development of BRIC have focused on preliminary feasibility testing using a laboratory- assembled pre-prototype unit. Observations of the drying behavior of actual brine solutions processed under BRIC-like conditions has been of particular interest. To date, experiments conducted with three types of analogue spacecraft wastewater brines have confirmed the basic premise behind the proposed application of in-place drying for these solutions. Specifically, the dried residual mass from these solutions have tended to exhibit characteristics of adhesion and flow that are expected to continue to challenge process stream management in spacecraft brine dewatering system designs. Yet, these same characteristics may favor the development of capillary- and surface-tension-based approaches envisioned as part of an ultimate microgravity-compatible BRIC design. In addition, preliminary feasibility testing of the BRIC pre-prototype confirmed that high rates of water recovery, up to 98% of the available brine water, may be possible while still removing the majority of the brine contaminants from the influent brine stream. These and other observations from testing are reported.

  6. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  7. The D&D of the Experimental Boiling Water Reactor (EBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Boling, L.E.; Yule, T.J.; Bhattacharyya, S.K.

    1996-03-01

    Argonne National Laboratory has completed the D&D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D&D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort.

  8. Photolytic treatment of atrazine-contaminated water: products, kinetics, and reactor design.

    Science.gov (United States)

    Ye, Xuejun; Chen, Daniel; Li, Kuyen; Wang, Bin; Hopper, Jack

    2007-08-01

    This study investigates the products, kinetics, and reactor design of atrazine photolysis under 254-nm ultraviolet-C (UVC) irradiation. With an initial atrazine concentration of 60 microg/L (60 ppbm), only two products remain in detectable levels. Up to 77% of decomposed atrazine becomes hydroxyatrazine, the major product. Both atrazine and hydroxyatrazine photodecompose following the first-order rate equation, but the hydroxyatrazine photodecomposition rate is significantly slower than that of atrazine. For atrazine photodecomposition, the rate constant is proportional to the square of UVC output, but inversely proportional to the reactor volume. For a photochemical reactor design, a series of equations are proposed to calculate the needed UVC output power, water treatment capacity, and atrazine outlet concentration.

  9. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW{sub th} and an electricity generation mode of 100 kW{sub th}, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  10. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Directory of Open Access Journals (Sweden)

    Seung Hyun Nam

    2015-10-01

    Full Text Available Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER, for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MWth and an electricity generation mode of 100 kWth, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  11. Drinking water treatment in solar reactors with immobilized photocatalysts

    Energy Technology Data Exchange (ETDEWEB)

    Sichel, C.; Fernandez, P.; Blanco, J.; Lorenz, K.

    2005-07-01

    This work has been performed at the Plataforma Solar de Almeria. As in our daily consumption of any other good, it is important to take an interest in sustainable treatment methods for purifying a vital water supply. Primary water treatment has no need for energy consuming techniques as any suspended particles can usually be removed by sand traps and sedimentation basin. Organic matter and biodegradable chemical contaminants ca be decomposed by activated sludge plants, bacteria beds, or in the case of highly organically loaded sewage by methanisation.In the recent years, another photocatalysts a photo sensitizer has been used in desinfection experiments. Ruthenium appears to have good potential for inactivation of bacteria in chelating coordination compounds. The SOLWATER project attempts to provide remote areas of such developing countries as Mexico, Peru and Argentina with drinking water disinfected by solar photocatalysis with immobilized TiO2 and Ru(II). (Author)

  12. Modern light water reactors : estate of the art, development and use of SWR1000; Reactores modernos de agua ligera. Situacion actual, posibilidades de desarrollo y empleo del SWR1000

    Energy Technology Data Exchange (ETDEWEB)

    Kaffine, H. W.

    2002-07-01

    Reactor vendors all over the world have been further developing various reactor design concepts, motivated by among other things, long-term energy prospects. One of the key areas in which Framatome ANP has been focusing such efforts is the further development of its boiling water reactor product line in the form of the SWR 1000. This plant design concept, developed on behalf of and in close collaboration with electric utilities and together with various European partners in accordance with the European Utility Requirements, features a level of nuclear safety that has been increased even further, particularly with respect to accident-induced releases of radioactivity, and power generating costs that are competitive with those of plants operating on other energy sources. Given the cost structure of nuclear power plants-in which capital charges account for around 60% of total generating cost, operating costs for some 24% and fuel cycle costs for approximately 16%-the main economic goal is to reduce the largest cost component: capital charges. In the case of the SWR 100, this is to be primarily achieved by means of an optimized, simplified plant design. The basic design of the SWR 1000, which was originally completed at the end of 1999 and has been undergoing further development in 2001, will soon be ready for commercial deployment. (Author)

  13. Laboratory data for review of outlet water temperature limits for BDF type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.; Fitzsimmons, D.E.

    1960-12-11

    A knowledge of the thermal and hydraulic conditions within a reactor fuel channel during an inadvertent flow reduction is needed to establish reactor operating limits. Such limits, which are based on outlet water temperature, are called ``trip-after-instability`` limits by the reactor operating personnel. Laboratory experiments were performed to update the knowledge of such conditions in a BDF reactor type fuel channel while using internally and externally cooled fuel elements (I&E`s) at tube powers up to 1530 KW. In addition to a general extension of previous data, the new data were used to review certain specific details involved in ``trip-after-instability`` limit calculations. It was found that in calculating the limits, the isothermal pressure drop across the fuel elements must be related to flow rate by the exponent 1.8, ({delta}P {proportional_to} F{sup 1.8}), rather than the more convenient value of 2.0. It was found that this method of limit determination is applicable to the high rear header pressures presently attained on the reactors and also applicable to tubes with very low Panellit pressures. And finally, the validity of certain analytical transformations of experimental data, called generalization of hydraulic demand curves, was reaffirmed for the above conditions.

  14. Microfluidic reactors for visible-light photocatalytic water purification assisted with thermolysis.

    Science.gov (United States)

    Wang, Ning; Tan, Furui; Wan, Li; Wu, Mengchun; Zhang, Xuming

    2014-09-01

    Photocatalytic water purification using visible light is under intense research in the hope to use sunlight efficiently, but the conventional bulk reactors are slow and complicated. This paper presents an integrated microfluidic planar reactor for visible-light photocatalysis with the merits of fine flow control, short reaction time, small sample volume, and long photocatalyst durability. One additional feature is that it enables one to use both the light and the heat energy of the light source simultaneously. The reactor consists of a BiVO4-coated glass as the substrate, a blank glass slide as the cover, and a UV-curable adhesive layer as the spacer and sealant. A blue light emitting diode panel (footprint 10 mm × 10 mm) is mounted on the microreactor to provide uniform irradiation over the whole reactor chamber, ensuring optimal utilization of the photons and easy adjustments of the light intensity and the reaction temperature. This microreactor may provide a versatile platform for studying the photocatalysis under combined conditions such as different temperatures, different light intensities, and different flow rates. Moreover, the microreactor demonstrates significant photodegradation with a reaction time of about 10 s, much shorter than typically a few hours using the bulk reactors, showing its potential as a rapid kit for characterization of photocatalyst performance.

  15. Single-stage temperature-controllable water gas shift reactor with catalytic nickel plates

    Science.gov (United States)

    Park, Jin-Woo; Lee, Sung-Wook; Lee, Chun-Boo; Park, Jong-Soo; Lee, Dong-Wook; Kim, Sung-Hyun; Kim, Sung-Soo; Ryi, Shin-Kun

    2014-02-01

    In this study, a microstructured reactor with catalytic nickel plates is newly designed and developed for proper heat management in an exothermic water gas shift WGS reaction. The reactor is designed to increase the reactor capacity simply by numbering-up a set of a catalyst layers and heat exchanger layers. The WGS reactor is built up with two sets of a catalyst layers and heat exchanger layers. The performance of the reactor is verified by WGS testing with the variation of the furnace temperatures, gas hourly space velocity (GHSV) and coolant (N2) flow rate. At a GHSV of 10,000 h-1, CO conversion reaches the equilibrium value with a CH4 selectivity of ≤0.5% at the furnace temperature of ≥375 °C. At high GHSV (40,000 h-1), CO conversion decreases considerably because of the heat from the exothermic WGS reaction at a large reactants mass. By increasing the coolant flow rate, the heat from the WGS reaction is properly managed, leading an increase of the CO conversion to the equilibrium value at GHSV of 40,000 h-1.

  16. Corrosion mechanisms of candidate structural materials for supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Lefu ZHANG; Fawen ZHU; Rui TANG

    2009-01-01

    Nickel-based alloys, austenitic stainless steel, ferritic/martensitic heat-resistant steels, and oxide dispersion strengthened steel are presently considered to be the candidate structural or fuel-cladding materials for supercritical water-cooled reactor (SCWR), one of the promising generation IV reactor for large-scale electric power production. However, corrosion and stress corrosion cracking of these candidate alloys still remain to be a major problem in the selection of nuclear fuel cladding and other structural materials, such as water rod. Survey of literature and experimental results reveal that the general corrosion mechanism of those candidate materials exhibits quite complicated mechanism in high-temperature and high-pressure supercritical water. Formation of a stable protective oxide film is the key to the best corrosion-resistant alloys. This paper focuses on the mechanism of corrosion oxide film breakdown for SCWR candidate materials.

  17. Environmentally assisted cracking in light water reactors annual report January - December 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

    2007-08-31

    This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

  18. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would

  19. Environmentally assisted cracking in Light Water Reactors. Volume 16: Semiannual report, October 1992--March 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289{degrees}C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy.

  20. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    Science.gov (United States)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  1. Gas phase photocatalytic water splitting in silicon based µ-reactors

    DEFF Research Database (Denmark)

    Dionigi, Fabio; Vesborg, Peter Christian Kjærgaard

    to the water splitting experiments, the results obtained with SrTiO2 and TiO2 are presented. These semiconductors are well known examples of materials active under UV illumination. However to achieve high efficiency of solar energy conversion the catalysts needs to be active for longer wavelength. Ga......N:ZnO is one of the few photocatalysts that is able to achieve overall water splitting with visible light. Therefore the reaction has been studied focusing on this material. GaN:ZnO loaded with Rh2-yCryO3 showed high activity and hydrogen and oxygen could even be detected under illumination with a solar light...... to the products detection using μ-reactors. In particular a new kind of μ-reactor that has a Pyrex lid on both sides is presented. With this reactor is possible to measure the absorbance of the materials deposited inside the μ-reactor and to combine optical measurements and spectroscopy with the detection...

  2. Biomimetic microchannels of planar reactors for optimized photocatalytic efficiency of water purification

    Science.gov (United States)

    Liao, Wuxia; Wang, Ning; Wang, Taisheng; Xu, Jia; Han, Xudong; Liu, Zhenyu; Yu, Weixing

    2016-01-01

    This paper reports a biomimetic design of microchannels in the planar reactors with the aim to optimize the photocatalytic efficiency of water purification. Inspired from biology, a bifurcated microchannel has been designed based on the Murray's law to connect to the reaction chamber for photocatalytic reaction. The microchannels are designed to have a constant depth of 50 μm but variable aspect ratios ranging from 0.015 to 0.125. To prove its effectiveness for photocatalytic water purification, the biomimetic planar reactors have been tested and compared with the non-biomimetic ones, showing an improvement of the degradation efficiency by 68%. By employing the finite element method, the flow process of the designed microchannel reactors has been simulated and analyzed. It is found that the biomimetic design owns a larger flow velocity fluctuation than that of the non-biomimetic one, which in turn results in a faster photocatalytic reaction speed. Such a biomimetic design paves the way for the design of more efficient planar reactors and may also find applications in other microfluidic systems that involve the use of microchannels. PMID:26958102

  3. A Study on the Optimal Position for the Secondary Neutron Source in Pressurized Water Reactors

    Directory of Open Access Journals (Sweden)

    Jungwon Sun

    2016-12-01

    Full Text Available This paper presents a new and efficient scheme to determine the optimal neutron source position in a model near-equilibrium pressurized water reactor, which is based on the OPR1000 Hanul Unit 3 Cycle 7 configuration. The proposed scheme particularly assigns importance of source positions according to the local adjoint flux distribution. In this research, detailed pin-by-pin reactor adjoint fluxes are determined by using the Monte Carlo KENO-VI code from solutions of the reactor homogeneous critical adjoint transport equations. The adjoint fluxes at each allowable source position are subsequently ranked to yield four candidate positions with the four highest adjoint fluxes. The study next simulates ex-core detector responses using the Monte Carlo MAVRIC code by assuming a neutron source is installed in one of the four candidate positions. The calculation is repeated for all positions. These detector responses are later converted into an inverse count rate ratio curve for each candidate source position. The study confirms that the optimal source position is the one with very high adjoint fluxes and detector responses, which is interestingly the original source position in the OPR1000 core, as it yields an inverse count rate ratio curve closest to the traditional 1/M line. The current work also clearly demonstrates that the proposed adjoint flux-based approach can be used to efficiently determine the optimal geometry for a neutron source and a detector in a modern pressurized water reactor core.

  4. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    Science.gov (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    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

  5. Inert matrix fuel neutronic, thermal-hydraulic, and transient behavior in a light water reactor

    Science.gov (United States)

    Carmack, W. J.; Todosow, M.; Meyer, M. K.; Pasamehmetoglu, K. O.

    2006-06-01

    Currently, commercial power reactors in the United States operate on a once-through or open cycle, with the spent nuclear fuel eventually destined for long-term storage in a geologic repository. Since the fissile and transuranic (TRU) elements in the spent nuclear fuel present a proliferation risk, limit the repository capacity, and are the major contributors to the long-term toxicity and dose from the repository, methods and systems are needed to reduce the amount of TRU that will eventually require long-term storage. An option to achieve a reduction in the amount, and modify the isotopic composition of TRU requiring geological disposal is 'burning' the TRU in commercial light water reactors (LWRs) and/or fast reactors. Fuel forms under consideration for TRU destruction in light water reactors (LWRs) include mixed-oxide (MOX), advanced mixed-oxide, and inert matrix fuels. Fertile-free inert matrix fuel (IMF) has been proposed for use in many forms and studied by several researchers. IMF offers several advantages relative to MOX, principally it provides a means for reducing the TRU in the fuel cycle by burning the fissile isotopes and transmuting the minor actinides while producing no new TRU elements from fertile isotopes. This paper will present and discuss the results of a four-bundle, neutronic, thermal-hydraulic, and transient analyses of proposed inert matrix materials in comparison with the results of similar analyses for reference UOX fuel bundles. The results of this work are to be used for screening purposes to identify the general feasibility of utilizing specific inert matrix fuel compositions in existing and future light water reactors. Compositions identified as feasible using the results of these analyses still require further detailed neutronic, thermal-hydraulic, and transient analysis study coupled with rigorous experimental testing and qualification.

  6. Reactor Materials Program process water piping indirect failure frequency

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W.L.

    1989-10-30

    Following completion of the probabilistic analyses, the LOCA Definition Project has been subject to various external reviews, and as a result the need for several revisions has arisen. This report updates and summarizes the indirect failure frequency analysis for the process water piping. In this report, a conservatism of the earlier analysis is removed, supporting lower failure frequency estimates. The analysis results are also reinterpreted in light of subsequent review comments.

  7. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, Progress Report for Work Through September 2002, 4th Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-09-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If no additional moderator is added to the fuel rod lattice, it is possible to attain fast neutron energy spectrum conditions in a supercritical water-cooled reactor (SCWR). This type of core can make use of either fertile or fertile-free fuel and retain a hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity. One can also add moderation and design a thermal spectrum SCWR. The Generation IV Roadmap effort has identified the thermal spectrum SCWR (followed by the fast spectrum SCWR) as one of the advanced concepts that should be developed for future use. Therefore, the work in this NERI project is addressing both types of SCWRs.

  8. Direct thermal water splitting by concentrated solar radiation for hydrogen production. Phase O: Proof of concept experiment

    Science.gov (United States)

    Genequand, P.

    1980-01-01

    The direct production of hydrogen from water and solar energy concentrated into a high temperature aperture is described. A solar powered reactor able to dissociate water vapor and to separate the reaction product at high temperature was developed, and direct water splitting has been achieved in a laboratory reactor. Water vapor and radiative heating from a carbon dioxide laser are fed into the reactor, and water vapor enriched in hydrogen and water vapor enriched in oxygen are produced. The enriched water vapors are separated through a separation membrane, a small disc of zirconium dioxide heated to a range of 1800 k to 2800 k. To avoid water vapor condensation within the reactor, the total pressure within the reactor was limited to 0.15 torr. A few modifications would enable the reactor to be operated at an increased pressure of a few torrs. More substantial modifications would allow for a reaction pressure of 0.1 atmosphere.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-11-15

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

  10. Efficiency Calibration of LaBr3(Ce) γ Spectroscopy in Analyzing Radionucles in Reactor Loop Water

    Institute of Scientific and Technical Information of China (English)

    CHEN; Xi-lin; QIN; Guo-xiu; GUO; Xiao-qing; CHEN; Yong-yong; MENG; Jun

    2013-01-01

    Monitoring the occurring and radioactivity concentration of fission products in nuclear reactor loop water is important for the nuclear reactor safe running evaluation,prevention of accidence and safe protection of working personnel.Study on the efficiency calibration for a LaBr3(Ce)detector experimental

  11. Boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant

    Energy Technology Data Exchange (ETDEWEB)

    Tokarev, Yu.I.; Sokolov, I.N.; Skvortsov, S.A.; Sidorov, A.M.; Krauze, L.V.

    1978-04-01

    The possibility of using a boiling water reactor in a prestressed reinforced concrete vessel for an atomic central heating-and-power plant (CHPP) was considered, with design features of the reactor intended for a two-purpose plant. A prestressed reinforced concrete vessel and integral arrangement of the primary circuit ensured reliability of the atomic CHPP using various CHPP flowsheets.

  12. Adapting computational optimization concepts from aeronautics to nuclear fusion reactor design

    Directory of Open Access Journals (Sweden)

    Baelmans M.

    2012-10-01

    Full Text Available Even on the most powerful supercomputers available today, computational nuclear fusion reactor divertor design is extremely CPU demanding, not least due to the large number of design variables and the hybrid micro-macro character of the flows. Therefore, automated design methods based on optimization can greatly assist current reactor design studies. Over the past decades, “adjoint methods” for shape optimization have proven their virtue in the field of aerodynamics. Applications include drag reduction for wing and wing-body configurations. Here we demonstrate that also for divertor design, these optimization methods have a large potential. Specifically, we apply the continuous adjoint method to the optimization of the divertor geometry in a 2D poloidal cross section of an axisymmetric tokamak device (as, e.g., JET and ITER, using a simplified model for the plasma edge. The design objective is to spread the target material heat load as much as possible by controlling the shape of the divertor, while maintaining the full helium ash removal capabilities of the vacuum pumping system.

  13. Concept of a nuclear powered submersible research vessel and a compact reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (Japan); Tokunaga, Sango [Japan Deep Sea Technology Association, Tokyo (Japan)

    2001-07-01

    A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

  14. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  15. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Science.gov (United States)

    Kulesza, Joel A.; Arzu Alpan, F.

    2016-02-01

    This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  16. Structural Integrity of Water Reactor Pressure Boundary Components.

    Science.gov (United States)

    1981-02-20

    RES-79-103 UNCLASSIFIED NRL--- 400 NURE-CR-17B3 NL mnmmnuunin -’El-.--. IIIIIIINI ., *q. - - ,aM T? * NUREG /CI 73 NIL Iteof AW, SOIituA 1 nert of Water...Progress Report for July-September 1979," NUREG /CR-1197, Oak Ridge National Labora- tory, Oak Ridge, Tn., Oct. 1978. 2. F. J. Loss, Ed., "Structural...Progress Report for April-June 1976," ORNL/ NUREG /TM-49, Oak Ridge National Labora- tory, Oak Ridge, Tn., Oct. 1976, pp. 27-38. 5. R. G. Berggren

  17. Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mohora, Emilijan, E-mail: emohora@ifc.org [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia); Roncevic, Srdjan; Dalmacija, Bozo; Agbaba, Jasmina; Watson, Malcolm; Karlovic, Elvira; Dalmacija, Milena [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A continuous electrocoagulation/flotation reactor was designed built and operated. Black-Right-Pointing-Pointer Highest NOM removal according to UV{sub 254} was 77% relative to raw groundwater. Black-Right-Pointing-Pointer Highest NOM removal accordance to DOC was 71%, relative to raw groundwater. Black-Right-Pointing-Pointer Highest As removal archived was 85% (6.2 {mu}g/l), relative to raw groundwater. Black-Right-Pointing-Pointer Specific reactor energy and electrode consumption was 1.7 kWh/m{sup 3} and 66 g Al/m{sup 3}. - Abstract: The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate = 4.3 l/h, inter electrode distance = 2.8 cm, current density = 5.78 mA/cm{sup 2}, A/V ratio = 0.248 cm{sup -1}. The NOM removal according to UV{sub 254} absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 {mu}g As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m{sup 3}. According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater.

  18. Co-digestion of concentrated black water and kitchen refuse in an accumulation system within the DESAR (decentralized sanitation and reuse) concept.

    Science.gov (United States)

    Kujawa-Roeleveld, K; Elmitwalli, T; Gaillard, A; van Leeuwen, M; Zeeman, G

    2003-01-01

    Co-digestion of concentrated black water and kitchen refuse within the DESAR concept was the objective of this pilot research. The digestion took place in two, non-mixed accumulation reactors (AC1 and AC2) inoculated with digested primary sludge from a WWTP at a temperature of 20 degrees C for a period of around 150 days. Reactor AC1 was fed with a mixture of faeces, urine and kitchen refuse in the equivalent amount that one individual generates per day. The AC2 was fed with a mixture of faeces and kitchen refuse in the equivalent amount that two individuals produce per day. Some contribution of urine to AC2 was not to be avoided. Detailed characterisation of waste(water) was performed. The performance of the stratified reactor was followed by monitoring the reactor content for several reactors' heights as well as being based on the biogas production. In general the system exposed good process stability. The methanisation of 34 and 61% was obtained for AC1 and AC2 respectively. The biogas yield was 26.5 and 50.8 L/p/d for the respective reactors. Proper choice of inoculum as well as good buffering capacity did not lead to accumulation of VFA and an inhibitive effect due to relatively high ammonium concentration. The chosen process is a promising technology showing good process stability especially for high strength influent.

  19. Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.; Gaitonde, S.M.

    1982-09-01

    The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

  20. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    Energy Technology Data Exchange (ETDEWEB)

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  1. A Flooding Induced Station Blackout Analysis for a Pressurized Water Reactor Using the RISMC Toolkit

    Directory of Open Access Journals (Sweden)

    Diego Mandelli

    2015-01-01

    Full Text Available In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. In addition, the impact of power uprate is determined in terms of both core damage probability and safety margins.

  2. International academic program in technologies of light-water nuclear reactors. Phases of development and implementation

    Science.gov (United States)

    Geraskin, N. I.; Glebov, V. B.

    2017-01-01

    The results of implementation of European educational projects CORONA and CORONA II dedicated to preserving and further developing nuclear knowledge and competencies in the area of technologies of light-water nuclear reactors are analyzed. Present article addresses issues of design and implementation of the program for specialized training in the branch of technologies of light-water nuclear reactors. The systematic approach has been used to construct the program for students of nuclear specialties, which corresponding to IAEA standards and commonly accepted nuclear principles recognized in the European Union. Possibilities of further development of the international cooperation between countries and educational institutions are analyzed. Special attention is paid to e-learning/distance training, nuclear knowledge preservation and interaction with European Nuclear Education Network.

  3. Removal of dissolved oxygen from water using a Pd-resin based catalytic reactor

    Institute of Scientific and Technical Information of China (English)

    Wenxin SHI; Chongwei CUI; Liye ZHAO; Shuili YU; Xia YUN

    2009-01-01

    The removal of dissolved oxygen (DO) from water was studied experimentally in a Pd-resin base catalyst reactor using purified hydrogen gas as a reducing agent. The effects of various operating conditions, such as hydrogen and water flow rates, height of the catalytic resin bed, temperature, pH value and nan time, on the removal of DO, had been studied extensively. The results shows that DO could be removed by the reactor from ppm to ppb levels at ambient temperature. Increases of temperature, H2gas rate and the height of the catalytic resin were helpful to improve the DO removal rate. The change of pH value fom 4 to 12 resulted in no effect on DO removal. Reaction time was the key factor to control the DO removal efficiency. Only when the reaction time was longer than 2.3 minutes under the experimental conditions, could a very low DO level be achieved.

  4. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  5. Flow-induced vibration for light water reactors. Final progress report, July 1981-September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Torres, M.R.

    1981-11-01

    Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, and general scaling laws to improve the accuracy of reduced-scale tests, and through the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976. This progress report summarizes the accomplishments achieved during the final period from July 1981 to September 1981. This is the last quarterly progress report to be issued for this program.

  6. Molecular Dynamics Simulations of Aqueous and Confined Systems Relevant to the Supercritical Water Cooled Nuclear Reactor

    Science.gov (United States)

    Kallikragas, Dimitrios Theofanis

    Supercritical water (SCW) is the intended heat transfer fluid and potential neutron moderator in the proposed GEN-IV Supercritical Water Cooled Reactor (SCWR). The oxidative environment poses challenges in choosing appropriate design materials, and the behaviour of SCW within crevices of the passivation layer is needed for developing a corrosion control strategy to minimize corrosion. Molecular Dynamics simulations have been employed to obtain diffusion coefficients, coordination number and surface density characteristics, of water and chloride in nanometer-spaced iron hydroxide surfaces. Diffusion models for hydrazine are evaluated along with hydration data. Results demonstrate that water is more likely to accumulate on the surface at low density conditions. The effect of confinement on the water structure diminishes as the gap size increases. The diffusion coefficient of chloride decreases with larger surface spacing. Clustering of water at the surface implies that the SCWR will be most susceptible to pitting corrosion and stress corrosion cracking.

  7. Design of an Organic Simplified Nuclear Reactor

    OpenAIRE

    Koroush Shirvan; Eric Forrest

    2016-01-01

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attr...

  8. Validation efforts for the neutronics of a plutonium erbium zirconium oxide inert matrix light water reactor fuel

    Science.gov (United States)

    Paratte, J. M.; Chawla, R.; Früh, R.; Joneja, O. P.; Pelloni, S.; Pralong, C.

    1999-08-01

    Light water reactor (LWR) neutronics codes and cross-section libraries need further qualification when used for the calculation of inert matrix fuel (IMF) cells. Three types of validation efforts have been undertaken for the PuO 2-Er 2O 3-ZrO 2 IMF concept under development at the Paul Scherrer Institute (PSI). Firstly, the PSI calculational scheme, based on the BOXER code and its data library, has been applied to the analysis of a range of LWR experiments with PuO 2-UO 2 fuel, conducted earlier at PSI's PROTEUS facility. The generally good agreement obtained between calculated and measured parameters gives confidence in the ability of the employed calculational scheme to correctly modelize Pu-containing fuel cells. Secondly, reactivity effects of various burnable poisons in a ZrO 2 matrix were measured in the CROCUS reactor of the Swiss Federal Institute of Technology at Lausanne. Modelling these experiments with BOXER resulted in satisfactory prediction of measured reactivity ratios (relative to a soluble-boron standard) for most of the experimental rods employed. This was particularly the case for experiments with erbium, as well as with mixtures of erbium and europium (the latter being used to simulate the effects of overlapping resonances, as would be expected in the case of a Pu-Er IMF). Finally, as there are no experimental results available from power reactors employing IMFs, the validation of burnup calculations (at the cell level) has been based on results obtained in the framework of an international benchmark exercise on the physics of LWRs employing IMFs. Certain discrepancies in calculated parameters have been observed in this context, several of which can be attributed to specific differences in cross-section libraries.

  9. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  10. Computational fluid dynamic analysis of a closure head penetration in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, D.R.; Schwirian, R.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-09-01

    ALLOY 600 has been used typically for penetrations through the closure head in pressurized water reactors because of its thermal compatibility with carbon steel, superior resistance to chloride attack and higher strength than the austenitic stainless steels. Recent plant operating experience with this alloy has indicated that this material may be susceptible to degradation. One of the major parameters relating to degradation of the head penetrations are the operational temperatures and stress levels in the penetration.

  11. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  12. The pressurization transient analysis for Lungmen advanced boiling water reactor using RETRAN-02

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, C.-W., E-mail: d937121@oz.nthu.edu.t [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Shih Chunkuan [Department of Engineering and System Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Institute of Nuclear Engineering and Science, National Tsing Hua University, No. 101, Sec. 2, Kuang Fu Road, Hsinchu 30013, Taiwan (China); Wang, J.-R.; Lin, H.-T. [Institute of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Cheng, S.-C. [Department of Nuclear Engineering, Taiwan Power Company, No. 242, Sec. 3, Roosevelt Rd., Taipei City 10016, Taiwan (China)

    2010-10-15

    A RETRAN-02 model was devised and benchmarked against the preliminary safety analysis report (PSAR) for the Lungmen nuclear power plant roughly 10 years ago. During these years, the fuel design, some of the reactor vessel designs, and control systems have since been revised. The Lungmen RETRAN-02 model has also been modified with updated information when available. This study uses the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen RETRAN-02 plant model. Five transients, load rejection (LR), turbine trip (TT), main steam line isolation valves closure (MSIVC), loss of feedwater flow (LOFF), and one turbine control valve closure (OTCVC), were utilized to validate the Lungmen RETRAN-02 model. Moreover, due to the strong coupling effect between neutron dynamics and the thermal-hydraulic response during pressurization of transients, the one-dimensional kinetic model with the cross-section data library is used to simulate the coupling effect. The analytical results show good agreement in trends between the RETRAN-02 calculation and the Lungmen FSAR data. Based on the benchmark of these design-basis transients, the modified Lungmen RETRAN-02 model has been adjusted to a level of confidence for analysis of pressure increase transients. Analytical results indicate that the Lungmen advanced boiling water reactor (ABWR) design satisfied design criteria, i.e., vessel pressure and hot shutdown capability. However, a slight difference exists in the simulation of the water level for cases with changes in water levels. The Lungmen RETRAN-02 model tends to predict the change in water level at a slower rate than that in the Lungmen FSAR. There is also a slight difference in void reactivity response toward vessel pressure change in both simulations, which causes the calculated neutron flux before reactor shutdown to differ to some degree when the reactor experiences a rapid pressure increase. Further studies will be performed in the future using

  13. Risk assessment of a pressurized water reactor for Class 3-8 accidents

    Energy Technology Data Exchange (ETDEWEB)

    Hall, R.E.

    1979-10-01

    An assessment has been made of the impact on societal risk of Class 3-8 accident sequences as defined by Appendix D to 10 CFR50. The present analysis concentrates on a pressurized water reactor and utilizes realistic assumptions when practical. Conclusions are drawn as to the relative importance of the analyzed accidents and their impact on the development of a complete societal risk curve. 65 refs., 61 figs., 37 tabs.

  14. WATER-GAS SHIFT KINETICS OVER IRON OXIDE CATALYSTS AT MEMBRANE REACTOR CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Carl R.F. Lund

    2002-08-02

    The kinetics of water-gas shift were studied over ferrochrome catalysts under conditions with high carbon dioxide partial pressures, such as would be expected in a membrane reactor. The catalyst activity is inhibited by increasing carbon dioxide partial pressure. A microkinetic model of the reaction kinetics was developed. The model indicated that catalyst performance could be improved by decreasing the strength of surface oxygen bonds. Literature data indicated that adding either ceria or copper to the catalyst as a promoter might impart this desired effect. Ceria-promoted ferrochrome catalysts did not perform any better than unpromoted catalyst at the conditions tested, but copper-promoted ferrochrome catalysts did offer an improvement over the base ferrochrome material. A different class of water-gas shift catalyst, sulfided CoMo/Al{sub 2}O{sub 3} is not affected by carbon dioxide and may be a good alternative to the ferrochrome system, provided other constraints, notably the requisite sulfur level and maximum temperature, are not too limiting. A model was developed for an adiabatic, high-temperature water-gas shift membrane reactor. Simulation results indicate that an excess of steam in the feed (three moles of water per mole of CO) is beneficial even in a membrane reactor as it reduces the rate of adiabatic temperature rise. The simulations also indicate that much greater improvement can be attained by improving the catalyst as opposed to improving the membrane. Further, eliminating the inhibition by carbon dioxide will have a greater impact than will increasing the catalyst activity (assuming inhibition is still operative). Follow-up research into the use of sulfide catalysts with continued kinetic and reactor modeling is suggested.

  15. Scram simulation of a control rod drive mechanism of a pressurized water reactor under seismic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, Katsuhisa; Shinohara, Yoshikazu; Ichinoo, Hiroyuki; Yoshikawa, Eiji; Nambu, Kiyoshi; Nomura, Tomonori.

    1987-03-01

    Control rod drop verification experiments of Mitsubishi pressurized water reactor under seismic conditions are performed to confirm the insertion function of control rods into the core. To evaluate these tests, computer simulations are performed. The scram time of control rods under seismic conditions was confirmed to meet the scram function. The behavior of the dropping control rods and the scram time obtained by the computer simulation show a very good correspondence with the results of verification experiments.

  16. On the factors influencing the performance of solar reactors for water disinfection with photosensitized singlet oxygen.

    Science.gov (United States)

    Manjón, Francisco; Villén, Laura; García-Fresnadillo, David; Orellana, Guillermo

    2008-01-01

    Two solar reactors based on compound parabolic collectors (CPCs) were optimized for water disinfection by photosensitized singlet oxygen (1O2) production in the heterogeneous phase. Sensitizing materials containing Ru(II) complexes immobilized on porous silicone were produced, photochemically characterized, and successfully tested for the inactivation of up to 10(4) CFU mL(-1) of waterborne Escherichia coli (gram-negative) or Enterococcus faecalis (gram-positive) bacteria. The main factors determining the performance of the solar reactors are the type of photosensitizing material, the sensitizer loading, the CPC collector geometry (fin- vs coaxial-type), the fluid rheology, and the balance between concurrent photothermal--photolytic and 1O2 effects on the microorganisms' inactivation. In this way, at the 40 degrees N latitude of Spain, water can be disinfected on a sunny day (0.6-0.8 MJ m(-2) L(-1) accumulated solar radiation dose in the 360-700 nm range, typically 5-6 h of sunlight) with a fin-type reactor containing 0.6 m2 of photosensitizing material saturated with tris(4,7-diphenyl-1,10-phenanthroline)ruthenium(II) (ca. 2.0 g m(-2)). The optimum rheological conditions require laminar-to-transitional water flow in both prototypes. The fin-type system showed better inactivation efficiency than the coaxial reactor due to a more important photolytic contribution. The durability of the sensitizing materials was tested and the operational lifetime of the photocatalyst is at least three months without any reduction in the bacteria inactivation efficiency. Solar water disinfection with 1O2-generating films is demonstrated to be an effective technique for use in isolated regions of developing countries with high yearly average sunshine.

  17. Modeling relations between the composition and properties of French light water reactor waste containment glass

    Energy Technology Data Exchange (ETDEWEB)

    Ghaleb, D.; Dussossoy, J.L.; Fillet, C.; Pacaud, F.; Jacquet-Francillon, N.

    1994-12-31

    Models have been developed to calculate the density, molten-state viscosity and initial corrosion rate according to the chemical composition of glass formulations used to vitrify high-level fission product solutions from reprocessed light water reactor fuel. Developed from other published work, these models have been adapted to allow for the effects of platinoid (Ru, Pd, Rh) inclusions on the molten glass rheology. (authors). 15 refs., 10 figs., 1 tab.

  18. Establishment of a Hub for the Light Water Reactor Sustainability Online Monitoring Community

    Energy Technology Data Exchange (ETDEWEB)

    Nancy J. Lybeck; Magdy S. Tawfik; Binh T. Pham

    2011-08-01

    Implementation of online monitoring and prognostics in existing U.S. nuclear power plants will involve coordinating the efforts of national laboratories, utilities, universities, and private companies. Internet-based collaborative work environments provide necessary communication tools to facilitate interaction between geographically diverse participants. Available technologies were considered, and a collaborative workspace was established at INL as a hub for the light water reactor sustainability online monitoring community.

  19. Metrics for the Evaluation of Light Water Reactor Accident Tolerant Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton

    2001-09-01

    The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of LWRs became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal of accident tolerant fuel (ATF) development is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. The complex multiphysics behavior of LWR nuclear fuel in the integrated reactor system makes defining specific material or design improvements difficult; as such, establishing desirable performance attributes is critical in guiding the design and development of fuels and cladding with enhanced accident tolerance. The U.S. Department of Energy is sponsoring multiple teams to develop ATF concepts within multiple national laboratories, universities, and the nuclear industry. Concepts under investigation offer both evolutionary and revolutionary changes to the current nuclear fuel system. This paper summarizes technical evaluation methodology proposed in the U.S. to aid in the optimization and down-selection of candidate ATF designs. This methodology will continue to be refined via input from the research community and industry, such that it is available to support the planned down-selection of ATF concepts in 2016.

  20. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor

    Energy Technology Data Exchange (ETDEWEB)

    Essadki, A.H., E-mail: essadki@hotmail.com [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Gourich, B. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Vial, Ch. [Laboratoire de Genie Chimique et Biochimique, LGCB-UBP/ENSCCF, 24 avenue des Landais, BP 206, 63174 Aubiere Cedex (France); Delmas, H. [Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France); Bennajah, M. [Ecole Superieure de Technologie de Casablanca, BP 8012, Oasis, Casablanca (Morocco); Laboratoire de Genie Chimique, ENSIACET-INPT, 5 rue Paulin Talabot, 31106 Toulouse (France)

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15 min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12 mA/cm{sup 2}, but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H{sub 2} microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  1. Defluoridation of drinking water by electrocoagulation/electroflotation in a stirred tank reactor with a comparative performance to an external-loop airlift reactor.

    Science.gov (United States)

    Essadki, A H; Gourich, B; Vial, Ch; Delmas, H; Bennajah, M

    2009-09-15

    Defluoridation using batch electrocoagulation/electroflotation (EC/EF) was carried out in two reactors for comparison purpose: a stirred tank reactor (STR) close to a conventional EC cell and an external-loop airlift reactor (ELAR) that was recently described as an innovative reactor for EC. The respective influences of current density, initial concentration and initial pH on the efficiency of defluoridation were investigated. The same trends were observed in both reactors, but the efficiency was higher in the STR at the beginning of the electrolysis, whereas similar values were usually achieved after 15min operation. The influence of the initial pH was explained using the analyses of sludge composition and residual soluble aluminum species in the effluents, and it was related to the prevailing mechanisms of defluoridation. Fluoride removal and sludge reduction were both favored by an initial pH around 4, but this value required an additional pre-treatment for pH adjustment. Finally, electric energy consumption was similar in both reactors when current density was lower than 12mA/cm(2), but mixing and complete flotation of the pollutants were achieved without additional mechanical power in the ELAR, using only the overall liquid recirculation induced by H(2) microbubbles generated by water electrolysis, which makes subsequent treatments easier to carry out.

  2. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  3. Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K. [Argonne National Lab., IL (United States)

    1996-01-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289{degrees}C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials.

  4. Analysis of transients in advanced heavy water reactor using lumped parameter models

    Energy Technology Data Exchange (ETDEWEB)

    Manmohan Pandey; Venkata Ramana Eaga; Sankar Sastry, P. [Department of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India); Gupta, S.K.; Lele, H.G.; Chatterjee, B. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

    2005-07-01

    Full text of publication follows: Analysis of transients occurring in nuclear power plants, arising from the complex interplay between core neutronics and thermal-hydraulics, is important for their operation and safety. Numerical simulations of such transients can be carried out extensively at very low computational cost by using lumped parameter mathematical models. The Advanced Heavy Water Reactor (AHWR), being developed in India, is a vertical pressure tube type reactor cooled by boiling light water under natural circulation, using thorium as fuel and heavy water as moderator. In the present work, nonlinear and linear lumped parameter dynamic models for AHWR have been developed and validated with a distributed parameter model. The nonlinear lumped model is based on point reactor kinetics equations and one-dimensional homogeneous equilibrium model of two-phase flow. The distributed model is built with RELAP5/MOD3.2 code. Various types of transients have been simulated numerically, using the lumped model as well as RELAP5. The results have been compared and parameters tuned to make the lumped model match the distributed model (RELAP5) in terms of steady state as well as dynamic behaviour. The linear model has been derived by linearizing the nonlinear model for small perturbations about the steady state. Numerical simulations of transients using the linear model have been compared with results obtained from the nonlinear model. Thus, the range of validity of the linear model has been determined. Stability characteristics of AHWR have been investigated using the lumped parameter models. (authors)

  5. Supercritical Water Reactor (SCWR) - Survey of Materials Research and Development Needs to Assess Viability

    Energy Technology Data Exchange (ETDEWEB)

    Philip E. MacDonald

    2003-09-01

    Supercritical water-cooled reactors (SCWRs) are among the most promising advanced nuclear systems because of their high thermal efficiency [i.e., about 45% vs. 33% of current light water reactors (LWRs)] and considerable plant simplification. SCWRs achieve this with superior thermodynamic conditions (i.e., high operating pressure and temperature), and by reducing the containment volume and eliminating the need for recirculation and jet pumps, pressurizer, steam generators, steam separators and dryers. The reference SCWR design in the U.S. is a direct cycle, thermal spectrum, light-water-cooled and moderated reactor with an operating pressure of 25 MPa and inlet/outlet coolant temperature of 280/500 °C. The inlet flow splits, partly to a down-comer and partly to a plenum at the top of the reactor pressure vessel to flow downward through the core in special water rods to the inlet plenum. This strategy is employed to provide good moderation at the top of the core, where the coolant density is only about 15-20% that of liquid water. The SCWR uses a power conversion cycle similar to that used in supercritical fossil-fired plants: high- intermediate- and low-pressure turbines are employed with one moisture-separator re-heater and up to eight feedwater heaters. The reference power is 3575 MWt, the net electric power is 1600 MWe and the thermal efficiency is 44.8%. The fuel is low-enriched uranium oxide fuel and the plant is designed primarily for base load operation. The purpose of this report is to survey existing materials for fossil, fission and fusion applications and identify the materials research and development needed to establish the SCWR viabilitya with regard to possible materials of construction. The two most significant materials related factors in going from the current LWR designs to the SCWR are the increase in outlet coolant temperature from 300 to 500 °C and the possible compatibility issues associated with the supercritical water environment.

  6. Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Junli Gou

    2009-01-01

    Full Text Available A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS, which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS, the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.

  7. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  8. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor

    Energy Technology Data Exchange (ETDEWEB)

    Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: vaveri@kist.re.kr; Kim, Jae-Duck [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: jdkim@kist.re.kr; Lee, Jong-Chol [Agency for Defense Development (ADD), P.O. Box 35-1, Yuseong-gu, Daejeon (Korea, Republic of)]. E-mail: jcleeadd@hanafos.com

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG (HOC{sub 2}H{sub 4}){sub 2}S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  9. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor.

    Science.gov (United States)

    Veriansyah, Bambang; Kim, Jae-Duck; Lee, Jong-Chol

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG, (HOC(2)H(4))2S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

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

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

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

  11. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  12. Solar radiation disinfection of drinking water at temperate latitudes: inactivation rates for an optimised reactor configuration.

    Science.gov (United States)

    Davies, C M; Roser, D J; Feitz, A J; Ashbolt, N J

    2009-02-01

    Solar radiation-driven inactivation of bacteria, virus and protozoan pathogen models was quantified in simulated drinking water at a temperate latitude (34 degrees S). The water was seeded with Enterococcus faecalis, Clostridium sporogenes spores, and P22 bacteriophage, each at ca 1x10(5) mL(-1), and exposed to natural sunlight in 30-L reaction vessels. Water temperature ranged from 17 to 39 degrees C during the experiments lasting up to 6h. Dark controls showed little inactivation and so it was concluded that the inactivation observed was primarily driven by non-thermal processes. The optimised reactor design achieved S90 values (cumulative exposure required for 90% reduction) for the test microorganisms in the range 0.63-1.82 MJ m(-2) of Global Solar Exposure (GSX) without the need for TiO2 as a catalyst. High turbidity (840-920 NTU) only reduced the S(90) value by 0.05). However, inactivation was significantly reduced for E. faecalis and P22 when the transmittance of UV wavelengths was attenuated by water with high colour (140 PtCo units) or a suboptimally transparent reactor lid (prob.waters and microorganisms. Although temperatures required for SODIS type pasteurization were not produced, non-thermal inactivation alone appeared to offer a viable means for reliably disinfecting low colour source waters by greater than 4 orders of magnitude on sunny days at 34 degrees S latitude.

  13. Flow injection spectrophotometry coupled with a crushed barium sulfate reactor column for the determination of sulfate ion in water samples

    OpenAIRE

    Burakham,Rodjana; Higuchi, Keiro; Oshima, Mitsuko; Grudpan, Kate; Motomizu, Shoji

    2004-01-01

    A new type of a reactor column, a crushed BaSO4 reactor column used for the flow injection spectrophotometric determination of sulfate ion using the exchange reaction of sulfate ion and barium-dimethylsulfonazo III is proposed. The column is very simple and economical. It can be continuously used for 8 h before washing with water for repeated usage of at least 1 month. The procedure is sensitive. Application to various water samples was demonstrated.

  14. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

  15. Radiological performance of hot water layer system in open pool type reactor

    Directory of Open Access Journals (Sweden)

    Amr Abdelhady

    2013-06-01

    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.

  16. Appendix to HDC 2118 design criteria 100-X reactor water plant, general description - section II

    Energy Technology Data Exchange (ETDEWEB)

    1952-03-29

    The factors responsible for the advances of 100-X compared with the older areas are: Simplification of the process, such as elimination of separate process water clearwells, by having the filtered water reservoirs perform that function. Combination of separate buildings into one building, such as combining filter pump house and process pump house. Use of electric standby. Use of higher capacity pumps and filter basins, and so fewer number of units. Centralization of control and operation. More compact arrangement of plant components. Use of waste heat for space heating, recovered from reactor effluent, backed up by steam plant.

  17. Membrane contactor/separator for an advanced ozone membrane reactor for treatment of recalcitrant organic pollutants in water

    Energy Technology Data Exchange (ETDEWEB)

    Chan, Wai Kit, E-mail: kekyeung@ust.hk [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Joueet, Justine; Heng, Samuel; Yeung, King Lun [Department of Chemical and Biomolecular Engineering, Hong Kong University of Science and Technology, Clear Water Bay, Kowloon (Hong Kong); Schrotter, Jean-Christophe [Water Research Center of Veolia, Anjou Recherche, Chemin de la Digue, BP 76. 78603, Maisons Laffitte, Cedex (France)

    2012-05-15

    An advanced ozone membrane reactor that synergistically combines membrane distributor for ozone gas, membrane contactor for pollutant adsorption and reaction, and membrane separator for clean water production is described. The membrane reactor represents an order of magnitude improvement over traditional semibatch reactor design and is capable of complete conversion of recalcitrant endocrine disrupting compounds (EDCs) in water at less than three minutes residence time. Coating the membrane contactor with alumina and hydrotalcite (Mg/Al=3) adsorbs and traps the organics in the reaction zone resulting in 30% increase of total organic carbon (TOC) removal. Large surface area coating that diffuses surface charges from adsorbed polar organic molecules is preferred as it reduces membrane polarization that is detrimental to separation. - Graphical abstract: Advanced ozone membrane reactor synergistically combines membrane distributor for ozone, membrane contactor for sorption and reaction and membrane separator for clean water production to achieve an order of magnitude enhancement in treatment performance compared to traditional ozone reactor. Highlights: Black-Right-Pointing-Pointer Novel reactor using membranes for ozone distributor, reaction contactor and water separator. Black-Right-Pointing-Pointer Designed to achieve an order of magnitude enhancement over traditional reactor. Black-Right-Pointing-Pointer Al{sub 2}O{sub 3} and hydrotalcite coatings capture and trap pollutants giving additional 30% TOC removal. Black-Right-Pointing-Pointer High surface area coating prevents polarization and improves membrane separation and life.

  18. Experimental demonstration of the reverse flow catalytic membrane reactor concept for energy efficient syngas production. Part 1: Influence of operating conditions

    NARCIS (Netherlands)

    Smit, J.; Bekink, G.J.; Sint Annaland, van M.; Kuipers, J.A.M.

    2007-01-01

    In this contribution the technical feasibility of the reverse flow catalytic membrane reactor (RFCMR) concept with porous membranes for energy efficient syngas production is investigated. In earlier work an experimental proof of principle was already provided [Smit, J., Bekink, G.J., van Sint Annala

  19. The use of active learning strategies in the instruction of Reactor Physics concepts

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Michael A.

    2000-01-01

    Each of the Active Learning strategies employed to teach Reactor Physics material has been or promises to be instructionally successful. The Cooperative Group strategy has demonstrated a statistically significant increase in student performance on the unit exam in teaching conceptually difficult, transport and diffusion theory material. However, this result was achieved at the expense of a modest increase in class time. The Tutorial CBI programs have enabled learning equally as well as classroom lectures without the direct intervention of an instructor. Thus, the Tutorials have been successful as homework assignments, releasing classroom time for other instruction. However, the time required for development of these tools was large, on the order of two hundred hours per hour of instruction. The initial introduction of the Case-Based strategy was roughly as effective as the traditional classroom instruction. Case-Based learning could well, after important modifications, perform better than traditional instruction. A larger percentage of the students prefer active learning strategies than prefer traditional lecture presentations. Student preferences for the active strategies were particularly strong when they believed that the strategies helped them learn the material better than they would have by using a lecture format. In some cases, students also preferred the active strategies because they were different from traditional instruction, a change of pace. Some students preferred lectures to CBI instruction, primarily because the CBI did not afford them the opportunity to question the instructor during the presentation.

  20. New bioproduction systems: from molecular circuits to novel reactor concepts in cell-free biotechnology.

    Science.gov (United States)

    Rupp, Steffen

    2013-01-01

    : The last decades witnessed a strong growth in several areas of biotechnology, especially in fields related to health, as well as in industrial biotechnology. Advances in molecular engineering now enable biotechnologists to design more efficient pathways in order to convert a larger spectrum of renewable resources into industrially used biofuels and chemicals as well as into new pharmaceuticals and therapeutic proteins. In addition material sciences advanced significantly making it more and more possible to integrate biology and engineering. One of the key questions currently is how to develop new ways of engineering biological systems to cope with the complexity and limitations given by the cell. The options to integrate biology with classical engineering advanced cell free technologies in the recent years significantly. Cell free protein production using cellular extracts is now a well-established universal technology for production of proteins derived from many organisms even at the milligram scale. Among other applications it has the potential to supply the demand for a multitude of enzymes and enzyme variants facilitating in vitro metabolic engineering. This review will briefly address the recent achievements and limitations of cell free conversions. Especially, the requirements for reactor systems in cell free biotechnology, a currently underdeveloped field, are reviewed and some perspectives are given on how material sciences and biotechnology might be able to advance these new developments in the future.

  1. New concept of designing Pu and MA containing fuel for fast reactors

    Science.gov (United States)

    Savchenko, A. M.; Konovalov, I. I.; Vatulin, A. V.; Glagovsky, E. M.

    2009-03-01

    New type of metal base fuel element is suggested for fast reactors. Basic approach to fuel element development - separated operations of fabricating uranium meat fuel element and introducing into it Pu or MA dioxides powder, that results in minimizing dust forming operations in fuel element fabrication. According to new fuel element design a framework fuel element having a porous uranium alloy meat is filled with standard PuO 2 powder of fuel meat metallurgically bonded to cladding forms a heat conducting framework, pores of which contain PuO 2 powder. Framework fuel element having porous meat is fabricated by capillary impregnation method with the use of Zr eutectic matrix alloys, which provides metallurgical bond between fuel and cladding and protects it from interaction. As compared to MOX fuel the new one features high thermal conductivity, higher uranium content, hence, high conversion ratio does not interact with fuel cladding and is more environmentally clean. Its principle advantage is a simple production process that is easily realized remotely, feasibility of involving high background Pu and MA isotopes into closed nuclear fuel cycle at the minimal influence on environment.

  2. Kinetic investigation of the oxidation of naval excess hazardous materials in supercritical water for the design of a transpiration-wall reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rice, S.F.; Hanush, R.G.; Hunter, T.B. [and others

    1997-01-01

    Experiments were conducted in Sandia`s supercritical fluids reactor (SFR) to generate data for the design of a transpiration-wall supercritical water oxidation (SCWO) reactor. The reactor is intended for the disposal of hazardous material generated on naval vessels. The design parameters for the system require an accurate knowledge of destruction efficiency vs. time and temperature. Three candidate materials were selected for testing. The experiments consisted of oxidizing these materials in the SFR at isothermal conditions over the temperature range of 400-550C at 24.1 MPa. A small extrapolation of the results shows that these materials can be adequately destroyed (to 99.9% destruction removal efficiency, DRE, based on total organic carbon (TOC) in the effluent) in approximately 5 seconds at 600C. The results vary smoothly and predictably with temperature such that extrapolation to higher temperatures beyond the experimental capabilities of the SFR can be made with reasonable confidence. The preliminary design of the transpiration-wall reactor has a rapid heat-up section within the reactor vessel that requires the addition of a fuel capable of quickly reacting with oxygen at temperatures below 500C. Candidate alcohols and JP-5 jet fuel were evaluated in this context. Oxidation rates for the alcohols were examined using in situ Raman spectroscopy. In addition, the potential utility of supplying the oxidizer line with hydrogen peroxide as an additive to enhance rapid initiation of the feed at unusually low temperatures was investigated. Experiments were conducted in the Supercritical Constant Volume Reactor (SCVR) using hydrogen peroxide as the initial oxidizing species. The results show that this concept as a method of enhancing low temperature reactivity appears to fail because thermal decomposition of the hydrogen peroxide is more rapid than the fuel oxidation rate at low temperatures. 8 refs., 16 figs., 5 tabs.

  3. A modified firefly algorithm applied to the nuclear reload problem of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Iona Maghali Santos de; Schirru, Roberto, E-mail: ioliveira@con.ufrj.b, E-mail: schirru@lmp.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    The Nuclear Reactor Reload Problem (NRRP) is an issue of great importance and concern in nuclear engineering. It is the problem related with the periodic operation of replacing part of the fuel of a nuclear reactor. Traditionally, this procedure occurs after a period of operation called a cycle, or whenever the nuclear power plant is unable to continue operating at its nominal power. Studied for more than 40 years, the NRRP still remains a challenge for many optimization techniques due to its multiple objectives concerning economics, safety and reactor physics calculations. Characteristics such as non-linearity, multimodality and high dimensionality also make the NRRP a very complex optimization problem. In broad terms, it aims at getting the best arrangement of fuel in the nuclear reactor core that leads to a maximization of the operating time. The primary goal is to design fuel loading patterns (LPs) so that the core produces the required energy output in an economical way, without violating safety limits. Since multiple feasible solutions can be obtained to this problem, judicious optimization is required in order to identify the most economical among them. In this sense, this paper presents a new contribution in this area and introduces a modified firefly algorithm (FA) to perform LPs optimization for a pressurized water reactor. Based on the original FA introduced by Xin-She Yang in 2008, the proposed methodology seems to be very promising as an optimizer to the NRRP. The experiments performed and the comparisons with some well known best performing algorithms from the literature, confirm this statement. (author)

  4. Grey water treatment concept integrating water and carbon recovery and removal of micropollutants

    NARCIS (Netherlands)

    Hernandez Leal, L.; Zeeman, G.; Buisman, C.J.N.

    2011-01-01

    A total treatment concept was developed for grey water from 32 houses in Sneek, The Netherlands. A thorough characterization of COD, nutrients, metals, micropollutants and anions was carried out. Four biological treatment systems were tested: aerobic, anaerobic, combined anaerobic¿+¿aerobic and a

  5. Grey water treatment concept integrating water and carbon recovery and removal of micropollutants

    NARCIS (Netherlands)

    Hernandez Leal, L.; Zeeman, G.; Buisman, C.J.N.

    2011-01-01

    A total treatment concept was developed for grey water from 32 houses in Sneek, The Netherlands. A thorough characterization of COD, nutrients, metals, micropollutants and anions was carried out. Four biological treatment systems were tested: aerobic, anaerobic, combined anaerobic¿+¿aerobic and a bi

  6. Analysis of the magnetic corrosion product deposits on a boiling water reactor cladding

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, Andrey [Paul Scherrer Institut, Villigen (Switzerland); Degueldre, Claude, E-mail: claude.degueldre@psi.ch [Paul Scherrer Institut, Villigen (Switzerland); Kaufmann, Wilfried [Kernkraftwerk Leibstadt, Leibstadt (Switzerland)

    2013-01-15

    The buildup of corrosion product deposits (CRUD) on the fuel cladding of the boiling water reactor (BWR) before and after zinc injection has been investigated by applying local experimental analytical techniques. Under the BWR water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}] spinel solid solutions. X-ray absorption spectroscopy (XAS) revealed inversion ratios of cation distribution in spinels deposited from the solid solution. Based on this information, a two-site ferrite spinel solid solution model is proposed. Electron probe microanalysis (EPMA) and extended X-ray absorption fine structure (EXAFS) findings suggest the zinc-rich ferrite spinels formation on BWR fuel cladding mainly at lower pin. - Graphical Abstract: Analysis of spinels in corrosion product deposits on boiling water reactor fuel rod. Combining EPMA and XAFS results: schematic representation of the ferrite spinels in terms of the end members and their extent of inversion. Note that the ferrites are represented as a surface between the normal (upper plane, M[Fe{sub 2}]O{sub 4}) and the inverse (lower plane, Fe[MFe]O{sub 4}). Actual compositions red Black-Small-Square for the specimen at low elevation (810 mm), blue Black-Small-Square for the specimen at mid elevation (1800 mm). The results have an impact on the properties of the CRUD material. Highlights: Black-Right-Pointing-Pointer Buildup of corrosion product deposits on fuel claddings of a boiling water reactor (BWR) are investigated. Black-Right-Pointing-Pointer Under BWR water conditions, Zn addition with Ni and Mn induced formation of (Zn,Ni,Mn)[Fe{sub 2}O{sub 4}]. Black-Right-Pointing-Pointer X-Ray Adsorption Spectroscopy (XAS) revealed inversion of cations in spinel solid solutions. Black-Right-Pointing-Pointer Zinc-rich ferrite spinels are formed on BWR fuel cladding mainly at lower pin elevations.

  7. Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seppaelae, Malla [VTT Technical Research Centre of Finland, P.O.Box 1000, FI02044 VTT (Finland)

    2008-07-01

    High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

  8. Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-08-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

  9. Degradation and detoxification of tar water from a gasification plant in a biogas reactor

    Energy Technology Data Exchange (ETDEWEB)

    Angelidaki, I.; Torry-Smith, M.; Petersen, L.; Ahring, B.K. [Denmarks Technical Univ., Dept. of Environmental Science and Engineering, Lyngby (Denmark)

    1998-08-01

    During gasification of biomass, toxic tar and tar-water (TW) are produced. The produced TW is saturated by soluble phenolic compound. In this project we have attempted detoxification of the produced TW by either wet oxidation where the TW is heated under pressure and addition of excess oxygen, and then subsequent treatment of the formed wet oxidized product in a biogas reactor. Alternatively, we treated the TW directly in a biogas reactor. Reduction in phenolic compounds, which constitute the major toxic compounds in TW, was used as a success parameter evaluating the detoxification of TW. The TW could successfully be degraded in a biogas reactor when co-digested with manure at a concentration up to 5%. Wet oxidized tar water (WOTW) could be degraded when added at a concentration of 30%. A biogas potential of approx. 0.19 L CH/g-VS was achieved for both the TW and WOTW. The biogas production per kg waste was 30 and 10 L CH/kg waste for the TW and the WOTW, respectively. The reason for the much lower methane production of the WOTW was that a large part of the organic content of the TW was converted to carbon dioxide during the wet oxidation process. The effluent concentrations of phenolic were in all cases much lower than the concentrations found in normally in undigested manure, showing that biomethanation of toxic wastes in co-digestion with manure could be an easy and cheap way to detoxify specific toxic wastes. (au)

  10. Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

    2017-05-15

    Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

  11. Storage of LWR (light-water-reactor) spent fuel in air

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, L.E.; Charlot, L.A.; Coleman, J.E. (Pacific Northwest Lab., Richland, WA (USA)); Knoll, R.W. (Johnson Controls, Inc., Madison, WI (USA))

    1989-12-01

    An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about