WorldWideScience

Sample records for lwr applications based

  1. An overview of advanced high-strength nickel-base alloys for LWR applications

    International Nuclear Information System (INIS)

    Prybylowski, J.; Ballinger, R.G.

    1989-01-01

    This paper reviews our current understanding of the behavior of high strength nickel base alloys used in light water reactor (LWR) applications. Emphasis is placed on understanding the fundamental mechanisms controlling crack propagation in these environments. To provide a foundation for this survey, general mechanisms of stress corrosion cracking and hydrogen embrittlement are first reviewed. The behavior of high strength nickel base alloys in LWR environments, as well as in other relevant environments is then reviewed. Suggested mechanisms of crack propagation are discussed. Alternate alloys and microstructural modifications that may result in improved behavior are presented. It is now clear that, at temperatures near 100C, alloy X-750, the predominant high strength nickel base alloy used today in LWR applications, is susceptible to hydrogen embrittlement. A review of published data from hydrogen embrittlement studies of nickel base superalloys during electrolytic charging and in hydrogen sulfide/brine solutions suggests that other nickel base superalloys are available possessing resistance to hydrogen embrittlement superior to that of alloy X-750. Available results of tests in gaseous hydrogen suggest that reduced grain boundary precipitation and a fine distribution of intragranular precipitates that act as irreversible hydrogen traps is the optimum microstructure for hydrogen embrittlement resistance. 42 refs., 2 figs., 5 tabs

  2. Implementation of static generalized perturbation theory for LWR design applications

    International Nuclear Information System (INIS)

    Byron, R.F.; White, J.R.

    1987-01-01

    A generalized perturbation theory (GPT) formulation is developed for application to light water reactor (LWR) design. The extensions made to standard generalized perturbation theory are the treatment of thermal-hydraulic and fission product poisoning feedbacks, and criticality reset. This formulation has been implemented into a standard LWR design code. The method is verified by comparing direct calculations with GPT calculations. Data are presented showing that feedback effects need to be considered when using GPT for LWR problems. Some specific potential applications of this theory to the field of LWR design are discussed

  3. Metal Matrix Microencapsulated Fuel Technology for LWR Applications

    International Nuclear Information System (INIS)

    Terrani, Kurt A.; Bell, Gary L.; Kiggans, Jim; Snead, Lance Lewis

    2012-01-01

    An overview of the metal matrix microencapsulated (M3) fuel concept for the specific LWR application has been provided. Basic fuel properties and characteristics that aim to improve operational reliability, enlarge performance envelope, and enhance safety margins under design-basis accident scenarios are summarized. Fabrication of M3 rodlets with various coated fuel particles over a temperature range of 800-1300 C is discussed. Results from preliminary irradiation testing of LWR M3 rodlets with surrogate coated fuel particles are also reported.

  4. Generalized perturbation theory for LWR depletion analysis and core design applications

    International Nuclear Information System (INIS)

    White, J.R.; Frank, B.R.

    1986-01-01

    A comprehensive time-dependent perturbation theory formulation that includes macroscopic depletion, thermal-hydraulic and poison feedback effects, and a criticality reset mechanism is developed. The methodology is compatible with most current LWR design codes. This new development allows GTP/DTP methods to be used quantitatively in a variety of realistic LWR physics applications that were not possible prior to this work. A GTP-based optimization technique for incore fuel management analyses is addressed as a promising application of the new formulation

  5. BUGLE-96: A revised multigroup cross section library for LWR applications based on ENDF/B-VI Release 3

    International Nuclear Information System (INIS)

    White, J.E.; Ingersoll, D.T.; Slater, C.O.; Roussin, R.W.

    1996-01-01

    A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs

  6. Characteristics Data Base: Programmer's guide to the LWR Quantities Data Base

    International Nuclear Information System (INIS)

    Jones, K.E.; Moore, R.S.

    1990-08-01

    The LWR Quantities Data Base is a menu-driven PC data base developed as part of OCRWM's waste, technical data base on the characteristics of potential repository wastes, which also includes non-LWR spent fuel, high-level and other materials. This programmer's guide completes the documentation for the LWR Quantities Data Base, the user's guide having been published previously. The PC data base itself may be requested from the Oak Ridge National Laboratory, using the order form provided in Volume 1 of publication DOE/RW-0184

  7. Selection methodology for LWR safety programs and proposals. Volume 2. Methodology application

    International Nuclear Information System (INIS)

    Ritzman, R.L.; Husseiny, A.A.

    1980-08-01

    The results of work done to update and apply a methodology for selecting (prioritizing) LWR safety technology R and D programs are described. The methodology is based on multiattribute utility (MAU) theory. Application of the methodology to rank-order a group of specific R and D programs included development of a complete set of attribute utility functions, specification of individual attribute scaling constants, and refinement and use of an interactive computer program (MAUP) to process decision-maker inputs and generate overall (multiattribute) program utility values. The output results from several decision-makers are examined for consistency and conclusions and recommendations regarding general use of the methodology are presented. 3 figures, 18 tables

  8. Fabrication of Multi-Layerd SiC Composite Tube for LWR Applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Daejong; Jung, Choonghwan; Kim, Weonju; Park, Jiyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Jongmin [Chungnam National Univ., Daejeon (Korea, Republic of)

    2013-05-15

    In this study, the chemical vapor deposition (CVD) and chemical vapor infiltration (CVI) methods were employed for the fabrication of the composite tubes. SiC ceramics and SiC-based composites have recently been studied for LWR fuel cladding applications because of good mechanical/physical properties, neutron irradiation resistance and excellent compatibility with coolant under severe accident. A multi-layered SiC composite tube as the nuclear fuel cladding is composed of the monolith SiC inner layer, SiC/SiC composite intermediate layer, and monolith SiC outer layer. Since all constituents should be highly pure, stoichiometric to achieve the good properties, it has been considered that the chemical process is a well-suited technique for the fabrication of the SiC phases.

  9. LWR reactivity/isotopics code for pedagogical and scoping applications

    International Nuclear Information System (INIS)

    AbuZaied, G.; Driscoll, M.J.

    1986-01-01

    A program designated BRICC (Burnup Reactivity and Isotopic Composition Computation), has been programmed for use on microcomputers to permit rapid parametric studies of the neutronics of light water reactor (LWR) assemblies. It is currently employed as a teaching tool in a graduate-level subject on nuclear fuel management, and has proven to be of sufficient accuracy to permit its use as a submodule in a more comprehensive program used to evaluate various mechanical spectral shift concepts for pressurized water reactor control. It should also prove useful in teaching reactor physics as it will fill an important gap between hand calculations of inadequate accuracy and state-of-the-art multigroup programs of daunting complexity. The BRICC program combines a minimum adequate set of old-fashioned phenomenological submodels that describe key physics attributed in an integral fashion, thereby providing the student or researcher with convenient mental pictures to serve as the basis for deductive reasoning. The program is short, written in a simplistic version of the Basic language, with many interspersed Remark statements, and is therefore easy to tinker with for various constructive purposes

  10. FABRICATION AND MATERIAL ISSUES FOR THE APPLICATION OF SiC COMPOSITES TO LWR FUEL CLADDING

    Directory of Open Access Journals (Sweden)

    WEON-JU KIM

    2013-08-01

    Full Text Available The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade SiCf/SiC composites are briefly reviewed. A CVI-processed SiCf/SiC composite with a PyC or (PyC-SiCn interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

  11. ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

    2000-11-01

    A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)

  12. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2B, User's guide to the LWR assemblies data base, Appendix 2C, User's guide to the LWR radiological data base, Appendix 2D, User's guide to the LWR quantities data base

    International Nuclear Information System (INIS)

    1987-12-01

    This User's Guide for the LWR Assemblies data base system is part of the Characteristics Data Base being developed under the Waste Systems Data Development Program. The objective of the LWR Assemblies data base is to provide access at the personal computer level to information about fuel assemblies used in light-water reactors. The information available is physical descriptions of intact fuel assemblies and radiological descriptions of spent fuel disassembly hardware. The LWR Assemblies data base is a user-oriented menu driven system. Each menu is instructive about its use. Section 5 of this guide provides a sample session with the data base to assist the user

  13. Conceptual development of a complete LWR reload design methodology based on generalized perturbation theory

    International Nuclear Information System (INIS)

    White, J.R.

    1986-01-01

    A new approach for the physics design and analysis of LWR reload cores is developed and demonstrated through several practical applications. The new design philosophy uses first- and second-order response derivatives to predict the important reactor performance characteristics (power peaking, reactivity coefficients, etc.) for any number of possible material configurations (assembly shuffling and burnable poison loadings). The response derivatives are computed using generalized perturbation theory (GPT) techniques. This report describes in detail an idealized GPT-based design system. The idealized system would contain individual modules to generate the required first-order and higher-order sensitivity data. It would also contain at least two major application codes; one for core design optimization and the other for evaluation of several safety parameters of interest in off-normal situations. This ideal system would be fully automated, user-friendly, and quite flexible in its ability to provide a variety of design and analysis capabilities. Information gained form these three studies gives a good foundation for the development of a complete integrated design package

  14. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    Directory of Open Access Journals (Sweden)

    Virpi Kouhia

    2012-01-01

    Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  15. PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

    International Nuclear Information System (INIS)

    Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

    2012-01-01

    This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

  16. Application of an enhanced cross-section interpolation model for highly poisoned LWR core calculations

    International Nuclear Information System (INIS)

    Palau, J.M.; Cathalau, S.; Hudelot, J.P.; Barran, F.; Bellanger, V.; Magnaud, C.; Moreau, F.

    2011-01-01

    Burnable poisons are extensively used by Light Water Reactor designers in order to preserve the fuel reactivity potential and increase the cycle length (without increasing the uranium enrichment). In the industrial two-steps (assembly 2D transport-core 3D diffusion) calculation schemes these heterogeneities yield to strong flux and cross-sections perturbations that have to be taken into account in the final 3D burn-up calculations. This paper presents the application of an enhanced cross-section interpolation model (implemented in the French CRONOS2 code) to LWR (highly poisoned) depleted core calculations. The principle is to use the absorbers (or actinide) concentrations as the new interpolation parameters instead of the standard local burnup/fluence parameters. It is shown by comparing the standard (burnup/fluence) and new (concentration) interpolation models and using the lattice transport code APOLLO2 as a numerical reference that reactivity and local reaction rate prediction of a 2x2 LWR assembly configuration (slab geometry) is significantly improved with the concentration interpolation model. Gains on reactivity and local power predictions (resp. more than 1000 pcm and 20 % discrepancy reduction compared to the reference APOLLO2 scheme) are obtained by using this model. In particular, when epithermal absorbers are inserted close to thermal poison the 'shadowing' ('screening') spectral effects occurring during control operations are much more correctly modeled by concentration parameters. Through this outstanding example it is highlighted that attention has to be paid to the choice of cross-section interpolation parameters (burnup 'indicator') in core calculations with few energy groups and variable geometries all along the irradiation cycle. Actually, this new model could be advantageously applied to steady-state and transient LWR heterogeneous core computational analysis dealing with strong spectral-history variations under

  17. Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

    International Nuclear Information System (INIS)

    Bond, W.D.; Mailen, J.C.; Michaels, G.E.

    1992-10-01

    The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO 2 fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed

  18. Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Bond, W.D.; Mailen, J.C.; Michaels, G.E.

    1992-10-01

    The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO[sub 2] fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

  19. Evaluation of methods for decladding LWR fuel for a pyroprocessing-based reprocessing plant

    Energy Technology Data Exchange (ETDEWEB)

    Bond, W.D.; Mailen, J.C.; Michaels, G.E.

    1992-10-01

    The first step in reprocessing disassembled light-water reactor (LWR) spent fuel is to separate the zirconium-based cladding from the UO{sub 2} fuel. A survey of decladding technologies has been performed to identify candidate decladding processes suitable for LWR fuel and compatible with downstream pyropr for separation of actinides and fission products. Technologies for the primary separation of Zircaloy cladding from oxide fuel and for secondary separations (in most cases, a further decontamination of the cladding) were reviewed. Because cutting of the fuel cladding is a necessary step in all flowsheet options, metal cutting technologies were also briefly evaluated. The assessment of decladding processes resulted in the identification of the three or four potentially attractive options that may warrant additional near-term evaluation. These options are summarized, and major strengths and issues of each option are discussed.

  20. Study on seismic base isolation of LWR plants

    International Nuclear Information System (INIS)

    Nakazawa, Masaaki; Nagano, Tooru; Kato, Asao; Kobatake, Masuhiko; Ohta, Kazuya.

    1991-01-01

    In general, laminated rubber bearings, while required to possess functions as base isolation devices, are also required to demonstrate functions as supportive structural materials for building superstructures. Accordingly, laminated rubber bearings are required to be durable for numbers of years about the same degrees as the service lives of the superstructures, and further, need to maintain their original functions. In this study, therefore, durability tests using small reduced models were performed to ascertain the durability of laminated natural rubber bearings. In evaluating durability of laminated rubber bearings, it is important to confirm how the functions as laminated rubber bearings (spring constant, breaking characteristics) will change with the environment and use over a long period of time. Therefore, experiments to confirm the items below were conducted for evaluation of the durability of laminated rubber bearings. 1) Influence of temperature of the environment in which laminated rubber bearings are used on the mechanical characteristics of the bearings; confirmation tests for temperature-dependent characteristics. 2) Influence of aging of rubber resulting from long-term use on mechanical characteristics (including breaking characteristics) of laminated rubber bearings; confirmation tests of time-dependent aging characteristics of laminated rubber bearings, excessive displacement loading test (including breaking tests) of laminated rubber bearings subjected to time-dependent aging. (J.P.N.)

  1. Development and application of best-estimate LWR safety analysis codes

    International Nuclear Information System (INIS)

    Reocreux, M.

    1997-01-01

    This paper is a review of the status and the future orientations of the development and application of best estimate LWR safety analysis codes. The present status of these codes exhibits a large success and almost a complete fulfillment of the objectives which were assigned in the 70s. The applications of Best Estimate codes are numerous and cover a large variety of safety questions. However these applications raised a number of problems. The first ones concern the need to have a better control of the quality of the results. This means requirements on code assessment and on uncertainties evaluation. The second ones concern needs for code development and specifically regarding physical models, numerics, coupling with other codes and programming. The analysis of the orientations for code developments and applications in the next years, shows that some developments should be made without delay in order to solve today questions whereas some others are more long term and should be tested for example in some pilot programmes before being eventually applied in main code development. Each of these development programmes are analyzed in the paper by detailing their main content and their possible interest. (author)

  2. Assessment of LWR piping design loading based on plant operating experience

    International Nuclear Information System (INIS)

    Svensson, P.O.

    1980-08-01

    The objective of this study has been to: (1) identify current Light Water Reactor (LWR) piping design load parameters, (2) identify significant actual LWR piping loads from plant operating experience, (3) perform a comparison of these two sets of data and determine the significance of any differences, and (4) make an evaluation of the load representation in current LWR piping design practice, in view of plant operating experience with respect to piping behavior and response to loading

  3. Stress corrosion cracking of age-hardenable nickel-base alloys in LWR-conditions

    International Nuclear Information System (INIS)

    Kekkonen, T.; Haenninen, H.

    1985-01-01

    At present it seems that the microstructure most resistant to stress corrosion cracking (SCC) in high temperature water is obtained by a solution annealing treatment at a relatively high temperature (appr. 1100 deg C) followed by water quenching and a single aging treatment (appr. 700 deg C/20 h). This should produce a microstructure with a high M 23 Cc 6 :MC ratio, semi-continous coherent M 23 C 6 precipitation, and an evenly distributed gamma prime in the matrix. However, since the actual mechanism of SCC in age-hardenable Ni-base alloys is unclear, the microstructural features resulting in the good resistance to SCC cannot be specified. Furthermore, the possible microstructural changes caused by prolonged use in LWR-conditions are unknown

  4. Some difference of concepts between design guideline for FBR base isolation system and aseismic design guideline of LWR in Japan

    International Nuclear Information System (INIS)

    Shibata, Heki

    1992-01-01

    This paper deals with the concept and the relation of 'the Base Isolation System and FBR' to the Safety Criteria and the Guideline of the Aseismic Design of LWR in Japan. The Central Research Institute of Electric Power Industries have been working for FBR last several years. The author has been contribute to their works, and this is one of the subjects. He described his own idea obtained through the cooperative work with CRIEPI. (author)

  5. Results of the LIRES Round Robin test on high temperature reference electrodes for LWR applications

    Energy Technology Data Exchange (ETDEWEB)

    Bosch, R.W. [SCK.CEN, Nuclear Research Centre Belgium, Boeretang 200, B-2400 Mol (Belgium); Nagy, G. [Magyar Tudomanyos Akademia KFKI Atomenergia Kutatointezet, AEKI, Konkoly Thege ut 29-33, 1121 Budapest (Hungary); Feron, D. [CEA Saclay, 91191 Gif-Sur-Yvette Cedex (France); Navas, M. [CIEMAT, Edificio 30, Dpto. Fision Nuclear, Avda. Complutense 22, 28040 Madrid, (Spain); Bogaerts, W. [KU Leuven, Kasteelpark Arenberg 31, B-3001 Leuven (Belgium); Karnik, D. [Nuclear Research Institute, NRI, Rez (Czech Republic); Dorsch, T. [Framatone ANP, Inc., Charlotte, North Carolina (United States); Molander, A. [Studsvik AB SE-611 82 Nykoeping (Sweden); Maekelae, K. [Materials and Structural Integrity, VTT Technical Research Centre of Finland, Kemistintie 3, P.O. Box 1704, FIN-02044 VTT (Finland)

    2004-07-01

    A European sponsored research project has been started on 1 October 2000 to develop high temperature reference electrodes that can be used for in-core electrochemical measurements in Light Water Reactors (LWR's). This LIRES-project (Development of Light Water Reactor Reference Electrodes) consists of 9 partners (SCK-CEN, AEKI, CEA, CIEMAT, KU Leuven, NRI Rez, Framatone ANP, Studsvik Nuclear and VTT) and will last for four years. The main objective of this LIRES project is to develop a reference electrode, which is robust enough to be used inside a LWR. Emphasize is put on the radiation hardness of both the mechanical design of the electrode as the proper functioning of the electrode. A four steps development trajectory is foreseen: (1) To set a testing standard for a Round Robin, (2) To develop different reference electrodes, (3) To perform a Round Robin test of these reference electrodes followed by selection of the best reference electrode(s), (4) To perform irradiation tests under appropriate LWR conditions in a Material Test Reactor (MTR). Four different high temperature reference electrodes have been developed and are being tested in a Round Robin test. These electrodes are: A Ceramic Membrane Electrode (CME), a Rhodium electrode, an external Ag/AgCl electrode and a Palladium electrode. The presentation will focus on the results obtained with the Round Robin test. (authors)

  6. Light Water Reactor (LWR) safety

    International Nuclear Information System (INIS)

    Sehgal, Bal Raj

    2006-01-01

    In this paper, a historical review of the developments in the safely of LWR power plants is presented. The paper reviews the developments prior to the TMI-2 accident, i.e. the concept of the defense in depth, the design basis, the large LOCA technical controversies and the LWR safety research programs. The TMI-2 accident, which became a turning point in the history of the development of nuclear power is described briefly. The Chernobyl accident, which terrified the world and almost completely curtailed the development of nuclear power is also described briefly. The great international effort of research in the LWR design-base and severe accidents, which was, respectively, conducted prior to and following the TMI-2 and Chernobyl accidents is described next. We conclude that with the knowledge gained and the improvements in plant organisation/management and in the training of the staff at the presently-installed nuclear power stations, the LWR plants have achieved very high standards of safety and performance. The Generation 3 + LWR power plants, next to be installed, may claim to have reached the goal of assuring the safety of the public to a very large extent. This review is based on the historical developments in LWR safety that occurred primarily in USA. however, they are valid for the rest of the Western World. This review can not do justice to the many many fine contributions that have been made over the last fifty years to the cause of LWR safety. We apologize if we have not mentioned them. We also apologize for not providing references to many of the fine investigations, which have contributed towards LWR safety earning the conclusions that we describe just above

  7. LWR-core behaviour project

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1982-07-01

    The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

  8. Study on material attractiveness aspect of spent nuclear fuel of LWR and FBR cycles based on isotopic plutonium production

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Saito, Masaki; Novitrian,; Waris, Abdul; Suud, Zaki

    2013-01-01

    Highlights: • The paper analyzes the plutonium production of recycling nuclear fuel option. • To evaluate material attractiveness based on intrinsic feature of material barrier. • Evaluation based on isotopic plutonium composition of spent fuel LWR and FBR. • Even mass number of plutonium gives a significant contribution to material barrier, in particular Pu-238 and Pu-240. • Doping MA in FBR blanket is effective to increase material barrier from weapon grade plutonium to more than MOX fuel grade. - Abstract: Recycling minor actinide (MA) as well as used uranium and plutonium can be considered to reduce nuclear waste production as well as to increase the intrinsic aspect of nuclear nonproliferation as doping material. Plutonium production as a significant aspect of recycling nuclear fuel option, gives some advantages and challenges, such as fissile material utilization of plutonium as well as production of some even mass number plutonium. The study intends to evaluate the material attractiveness based on the intrinsic feature of material barrier such as plutonium composition, decay heat and spontaneous fission neutron components from spent fuel (SF) light water reactor (LWR) and fast breeder reactor (FBR) cycles. A significant contribution has been shown by decay heat (DH) and spontaneous fission neutron (SFN) of even mass number of plutonium isotopes to the total DH and SFN of plutonium element, in particular from isotopic plutonium Pu-238 and Pu-240 contributions. Longer decay cooling time and higher burnup are effective to increase the material barrier (DH and SFN) level from reactor grade plutonium level to MOX grade plutonium level. Material barrier of plutonium element from spent fuel (SF) FBR in the core regions has similarity to the material barrier profile of SF LWR which can be categorized as MOX fuel grade plutonium. Plutonium compositions, DH and SFN components are categorized as weapon grade plutonium level for FBR blanket regions with no

  9. Application of the Combined Cycle LWR-Gas Turbine to PWR for NPP Life Extension Safety Upgrade and Improving Economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.

    2006-01-01

    Currently, some of the most important problem for the nuclear industry are life extension, advance competitiveness and safety of aging LWR NPPs. Based on results of studies performed in the USA (Battelle Memorial Institute) and in Russia (NIKIET), a new power technology, using a combined cycle gas-turbine facility CCGT - LWR, so called TD-Cycle, can significantly help in resolution of some problems of nuclear power industry. The nuclear steam and gas topping cycle is used for re-powering a light water pressurized reactor of PWR or VVER type. An existing NPP is topped with a gas turbine facility with a heat recovery steam generator (HRSG) generating steam from waste heat. The superheated steam of high pressure (P=90-165 bar, T=500-550 C) generated in the HRSG, is expanded in a high pressure (HP) turbine for producing electricity. The HP turbine can work on one shaft with the the gas turbine or at one shaft with intermediate (IP) or low (LP) pressure parts of the main nuclear steam turbine, or with a separate electric generator. The exhausted steam from the HP turbine is injected into the steam mixer where it is mixed with the saturated steam from the NPP steam generator (SG). The mixer is intended to superheat the main nuclear steam and should be characterized by minimum losses during mixing superheated and saturated steam. Steam from the mixer superheated by 20-60 C directs to the existing IP turbine, and then, through a separator-reheater flows into the LP turbine. Feed water re-heaters of LP and HP are actually unchanged in this case. Feed water extraction to the HRSG is supplied after one of LP water heaters. This proposal is intended to re-power existing LWR NPPs. To minimize cost, the IP and LP turbines and electric generator would remain the same. The reactor thermal power and fast neutron flux to the reactor vessel would decrease by 30-50 percent of nominal values. The external peripheral row of fuel elements can be replaced with metal absorber rods to

  10. Validation of LWR calculation methods and JEF-1 based data libraries by TRX and BAPL critical experiments

    International Nuclear Information System (INIS)

    Pelloni, S.; Grimm, P.; Mathews, D.; Paratte, J.M.

    1989-06-01

    In this report the capability of various code systems widely used at PSI (such as WIMS-D, BOXER, and the AARE modules TRAMIX and MICROX-2 in connection with the one-dimensional transport code ONEDANT) and JEF-1 based nuclear data libraries to compute LWR lattices is analysed by comparing results from thermal reactor benchmarks TRX and BAPL with experiment and with previously published values. It is shown that with the JEF-1 evaluation eigenvalues are generally well predicted within 8 mk (1 mk = 0.001) or less by all code systems, and that all methods give reasonable results for the measured reaction rate within or not too far from the experimental uncertainty. This is consistent with previous similar studies. (author) 7 tabs., 36 refs

  11. The evolutionary adoption of thorium beginning with its application in niche LWR fuels

    International Nuclear Information System (INIS)

    Drera, Saleem

    2015-01-01

    Since the inception of nuclear energy, the use of thorium as a nuclear fuel has been envisioned. Thorium boasts benefits, however, drawbacks which are both economic and technical including its the lack of a naturally occurring fissile isotope implies that its utility is inherently more difficult. The implementation of thorium as a nuclear fuel requires that it must provide sound technical advantages in combination with attractive economics as compared to standard uranium fuel. Revolutionary thorium concepts such as molten salt reactors and accelerator driven systems may provide theoretical merit, however, their exotic nature and associated technical challenges label them as long-term solutions at best. A near-to-medium term solution for thorium must be based on an evolutionary approach utilizing light/heavy water reactor platforms. While thorium does not provide a near-to-medium term complete replacement of uranium, it does provide substantial benefit within niche applications. To license and bring to market these niche fuels, Thor Energy and an international consortium of entities (including: Fortum, KAERI, Westinghouse, NNL, ITU, IFE, and a few other minor entities) have initiated a fuel development and irradiation test program to characterize the performance of these thoria-containing fuels. (author)

  12. Hydrogen permeation in FeCrAl alloys for LWR cladding application

    Science.gov (United States)

    Hu, Xunxiang; Terrani, Kurt A.; Wirth, Brian D.; Snead, Lance L.

    2015-06-01

    FeCrAl, an advanced oxidation-resistant iron-based alloy class, is a highly prevalent candidate as an accident-tolerant fuel cladding material. Compared with traditional zirconium alloy fuel cladding, increased tritium permeation through FeCrAl fuel cladding to the primary coolant is expected, raising potential safety concerns. In this study, the hydrogen permeability of several FeCrAl alloys was obtained using a static permeation test station, which was calibrated and validated using 304 stainless steel. The high hydrogen permeability of FeCrAl alloys leads to concerns with respect to potentially significant tritium release when used for fuel cladding in LWRs. The total tritium inventory inside the primary coolant of a light water reactor was quantified by applying a 1-dimensional steady state tritium diffusion model to demonstrate the dependence of tritium inventory on fuel cladding type. Furthermore, potential mitigation strategies for tritium release from FeCrAl fuel cladding were discussed and indicate the potential for application of an alumina layer on the inner clad surface to serve as a tritium barrier. More effort is required to develop a robust, economical mitigation strategy for tritium permeation in reactors using FeCrAl clad fuel assemblies.

  13. Summary remarks and recommended reactions for an international data file for dosimetry applications for LWR, FBR, and MFR reactor research, development and testing programs

    International Nuclear Information System (INIS)

    McElroy, W.N.; Lippincott, E.P.; Grundl, J.A.; Fabry, A.; Dierckx, R.; Farinelli, U.

    1979-01-01

    The need for the use of an internationally accepted data file for dosimetry applications for light water reactor (LWR), fast breeder reactor (FBR), and magnetic fusion reactor (MFR) research, development, and testing programs continues to exist for the Nuclear Industry. The work of this IAEA meeting, therefore, will be another important step in achieving consensus agreement on an internationally recommended file and its purpose, content, structure, selected reactions, and associated uncertainy files. Summary remarks and a listing of recommended reactions for consideration in the formulation of an ''International Data File for Dosimetry Applications'' are presented in subsequent sections of this report

  14. The applicability of detailed process for neutron resonance absorption to neutronics analyses in LWR next generation fuels to extend burnup

    International Nuclear Information System (INIS)

    Kameyama, Takanori; Nauchi, Yasushi

    2004-01-01

    Neutronics analyses with detail processing for neutron resonance absorption in LWR next generation UOX and MOX fuels to extend burnup were performed based on the neutronic transport and burnup calculation. In the detailed processing, ultra-fine energy nuclear library and collision probabilities between neutron and U, Pu nuclides (actinide nuclides) are utilized for two-dimension geometry. In the usual simple processing (narrow resonance approximation), shielding factors and compensation equations for neutron resonance absorption are utilized. The results with detailed and simple processing were compared to clarify where the detailed processing is needed. The two processing caused difference of neutron multiplication factor by 0.5% at the beginning of irradiation, while the difference became smaller as burnup increased and was not significant at high burnup. The nuclide compositions of the fuel rods for main actinide nuclides were little different besides Cm isotopes by the processing, since the neutron absorption rate of 244 Cm became different. The detail processing is needed to evaluate the neutron emission rate in spent fuels. In the fuel assemblies, the distributions of rod power rates were not different within 0.5%, and the peak rates of fuel rod were almost the same by the two processing at the beginning of irradiation when the peak rate is the largest during the irradiation. The simple processing is also satisfied for safety evaluation based on the peak rate of rod power. The difference of local power densities in fuel pellets became larger as burnup increased, since the neutron absorption rate of 238 U in the peripheral region of pellets were significantly different by the two processing. The detail processing is needed to evaluate the fuel behavior at high burnup. (author)

  15. Development and application of an LWR reactor pressure vessel-specific flaw distribution

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.

    1991-01-01

    Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses

  16. Status of LWR fuel design and future usage of JENDL

    International Nuclear Information System (INIS)

    Ito, Takuya

    2008-01-01

    For all conventional LWR fuel design codes of LWR fuel manufactures in Japan, the cross section library are based on the ENDF/B. Recently we can see several movements for the utilization of JENDL library for the LWR fuel design. The latest version of NEUPHYS cross section library is based on the JENDL-3.2. To accelerate this movement of JENDL utilization in LWR fuel design, it is necessary to prepare a high quality JENDL document, systematic validation of JENDL and to appeal them abroad effectively. (author)

  17. Creep damage in zircaloy-4 at LWR temperatures

    International Nuclear Information System (INIS)

    Keusseyan, R.L.; Hu, C.P.; Li, C.Y.

    1978-08-01

    The observation of creep damage in the form of grain boundary cavitation in Zircaloy-4 in the temperature range of interest to Light Water Reactor (LWR) applications is reported. The observed damage is shown to reduce the ductility of Zircaloy-4 in a tensile test at LWR temperatures

  18. LWR physics in SKODA Works

    International Nuclear Information System (INIS)

    Zbytovsky, A.; Lehmann, M.; Vyskocil, V.; Vacek, J.; Krysl, V.

    1980-01-01

    Computation of nuclear power reactors of the WWER-1000 type is described as are computer programs used by Skoda Works for the solution of neutron problems. The programs are analyzed for applicability in the unified program system of the CMEA countries which will be used in the preparation of safety reports, the evaluation of safety hazards, the design of fuel charges, economical studies etc. A detailed description is also presented of multigroup transport calculations and of the preparation of input data for macrocalculations of the heterogeneous lattices of LWR's. (author)

  19. The prediction of the LWR plant accident based on the measured plant data

    International Nuclear Information System (INIS)

    Miettinen, J.; Schmuck, P.

    2005-01-01

    In case of accident affecting a nuclear reactor, it is essential to anticipate the possible development of the situation to efficiently succeed in emergency response actions, i.e. firstly to be early warned, to get sufficient information on the plant: and as far as possible. The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) project consists in developing a methodology: of expertise to; structure the work of technical teams and to facilitate cross competence communications among EP players and a qualified computer tool that could be commonly used by the European countries to reliably predict source term in case of an accident in a light water reactor, using the information available on the plant. In many accident conditions the team of analysts may be located far away from the plant experiencing the accident and their decision making is based on the on-line plant data transmitted into the crisis centre in an interval of 30 - 600 seconds. The plant condition has to be diagnosed based on this information, In the ASTRID project the plant status diagnostics has been studied for the European reactor types including BWR, PWR and VVER plants. The directly measured plant data may be used for estimations of the break size from the primary system and its locations. The break size prediction may be based on the pressurizer level, reactor vessel level, primary pressure and steam generator level in the case of the steam generator tube rupture. In the ASTRID project the break predictions concept was developed and its validity for different plant types and is presented in the paper, when the plant data has been created with the plant specific thermohydraulic simulation model. The tracking simulator attempts to follow the plant behavior on-line based on the measured plant data for the main process parameters and most important boundary conditions. When the plant state tracking fails, the plant may be experiencing an accident, and the tracking

  20. Technical bases for estimating fission product behavior during LWR accidents. Technical report

    International Nuclear Information System (INIS)

    1981-06-01

    The objective of this report is to provide the Nuclear Regulatory Commission and the public with a description of the best technical information currently available for estimating the release of radioactive material during postulated reactor accidents, and to identify where gaps exist in our knowledge. This report focuses on those low probability-high consequence accidents involving severe damage to the reactor core and core meltdown that dominate the risk to the public. Furthermore, in this report particular emphasis is placed on the accident behavior of radioactive iodine, as (1) radioiodine is predicted to be a major contributor to public exposure, (2) current regulatory accident analysis procedures focus on iodine, and (3) several technical issues have been raised recently about the magnitude of iodine release. The generation, transport, and attenuation of aerosols were also investigated in some detail to assess their effect on fission product release estimates and to determine the performance of engineered safety features under accident conditions exceeding their design bases

  1. Is it the end of history for LWR safety?

    International Nuclear Information System (INIS)

    Sehgal, Bal Raj

    2004-01-01

    In this essay a parallel is drawn between the struggle for recognition, which is argued by Fukuyama as the 'motor' of human history and that waged by the LWR safety for the public to recognize the LWR plants as a source of safe nuclear power. The end of history for the ''human struggle for recognition'' as the capitalistic liberal democracy is equated with the ''end of history'' for the LWR safety to provide assurance to the public of termination of a severe accident it ever would occur. It is suggested that we are near ''the end of history'' of the LWR safety for the new-design LWR plants but fall short for the presently-installed plants. The essay bases these suggestions on an examination of the history of nuclear power development in U.S.A., but also considering the more recent regulatory and public acceptance developments in Europe and the rest of the World. (author)

  2. Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

    International Nuclear Information System (INIS)

    Chupin, A.; Hu, L. W.; Buongiorno, J.

    2008-01-01

    Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

  3. Thermo-mechanical assessment of full SiC/SiC composite cladding for LWR applications with sensitivity analysis

    Science.gov (United States)

    Singh, Gyanender; Terrani, Kurt; Katoh, Yutai

    2018-02-01

    SiC/SiC composites are considered among leading candidates for accident tolerant fuel cladding in light water reactors. However, when SiC-based materials are exposed to neutron irradiation, they experience significant changes in dimensions and physical properties. Under a large heat flux application (i.e. fuel cladding), the non-uniform changes in the dimensions and physical properties will lead to build-up of stresses in the structure over the course of time. To ensure reliable and safe operation of such a structure it is important to assess its thermo-mechanical performance under in-reactor conditions of irradiation and elevated temperature. In this work, the foundation for 3D thermo-mechanical analysis of SiC/SiC cladding is put in place and a set of analyses with simplified boundary conditions has been performed. The analyses were carried out with two different codes that were benchmarked against one another and prior results in the literature. A constitutive model is constructed and solved numerically to predict the stress distribution and variation in the cladding under normal operating conditions. The dependence of dimensions and physical properties variation with irradiation and temperature has been incorporated. These robust models may now be modified to take into account the axial and circumferential variation in neutron and heat flux to fully account for 3D effects. The results from the simple analyses show the development of high tensile stresses especially in the circumferential and axial directions at the inner region of the cladding. Based on the results obtained, design guidelines are recommended. For lack of certainty in or tailor-ability for the physical and mechanical properties of SiC/SiC composite material a sensitivity analysis is conducted. The analysis results establish a precedence order of the properties based on the extent to which these properties influence the temperature and the stresses.

  4. EURLIB-LWR-45/16 and - 15/5. Two board group libraries for LWR-shielding problems

    Energy Technology Data Exchange (ETDEWEB)

    Herrnberger, V

    1982-04-01

    Specifications of the broad group cross section libraries EURLIB-LWR-45/16 and -15/5 are given. They are based on EURLIB-III data and produced for LWR shielding problems. The elements considered are H, C{sub 12}, O, Na, Al, Si, Ca, Cr, Mn, Fe, Ni, Zr, U{sub 235}, U{sub 238}. The cross section libraries are available upon request from EIR, RSIC, NEA-CPL and IAEA-NDS. (author) Refs, figs, tabs

  5. 'CANDLE' burnup regime after LWR regime

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nagata, Akito

    2008-01-01

    CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

  6. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu.N.; Gabaraev, B.A.

    2002-01-01

    Full text: The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper - erection of steam-gas toppings to the nuclear power units - is considered in the paper. Application of the steam-gas toppings permits through reducing power of ageing reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project, for Russian boiling VK-50 reactor now in operation Application of the steam-gas topping permits: extend the service life of ageing VVER-440 reactor by 10...15 years; use the turbine and other NPP balance-of-plant equipment at full power; increase the efficiency of combined cycle up to 48% and more; enhance the safety of NPP operation; utilize NPP balance-of-plant equipment after reactor decommissioning; perform the cost-effective operation in maneuvering modes; increase capacity factor of the plant. The construction of pilot project on the basis of the VK-50 reactor will allow not only to demonstrate new technology but also to attain appreciable economic effect including that obtained due to using the available reserves of the NPP turbine. (author)

  7. Manufacturing of FeCrAl/Zr Dual Layer tube for its application to LWR Fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Dong Jun; Lim, Do Wan; Jung, Yang Il; Kim, Hyun Gil; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    Many advanced materials such as MAX phases, Mo, SiC, and Fe-based alloys are being considered a possible candidate to substitute the Zr-based alloy cladding has been used in light water reactors. Among the proposed candidate materials, Fe-based alloy is one of the most promising candidates owing to its excellent formability, very good high strength, and corrosion resistance at high temperature. However, neutron cross section of FeCrAl alloy is much higher than that of existing Zr-based alloys. In this study, FeCrAl/Zr dual layer tube was manufactured by using a hot isostatic pressing (HIP) method. The thickness of outer FeCrAl layer was varied from 50 to 250 μm but all the FeCrAl/Zr dual layer tube samples maintained its total thickness of 570 μm. For a detailed microstructural characterization of FeCrAl/Zr dual layer, polarized optical microscopy and scanning electron microscopy (SEM) study carried out and its mechanical property was measured by ring compression test. FeCrAl/Zr dual layer tube sample was successfully manufactured with good adhesion between both layers. Inter layer showing gradual element variation was observed at interface. Result obtained from simulated LOCA test indicates that FeCrAl/Zr dual layer tube may maintain its integrity during LOCA and its accident tolerance had greatly improved compared to that of Zr-based alloy.

  8. Application of the combined cycle LWR-gas turbine to PWR for NPP life extension, safety upgrade and improving economy

    International Nuclear Information System (INIS)

    Kuznetsov, Yu. N.; Lisitsa, F. D.; Smirnov, V. G.

    2000-01-01

    The unconventional technology to extend the lifetime for the NPPs now in operation and make a construction of new NPPs cheaper three-quarter erection of steam-gas toppings to the nuclear power units three-quarter is considered in the paper. Application of the steam-gas toppings permits through reducing power of aging reactors to extend lifetime of nuclear power unit, enhance its safety and at the same time to keep full load operation of NPP turbine and other balance-of-plant equipment. Proposed technology is examined for Russian VVER-440 reactor as an example and, also, as a pilot project. for Russian boiling VK-50 reactor now in operation. Heat flow sheets of the power plants, their parameters and economic problems are discussed. (author)

  9. Validation of the code ETOBOX/BOXER for UO2 LWR lattices based on the experiments TRX, BAPL-UO2 and other critical experiments

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1985-07-01

    The EIR codes system for LWR arrays is based on cross sections taken out of ENDF/B-4 and ENDF/B-5 by the code ETOBOX. The calculation method for the arrays (code BOXER) and the cross sections as well were applied to the CSEWG benchmark experiments TRX-1 to 4 and BAPL-UO/sub 2/-1 to 3. The results are compared to the measured values and to some calculations of other institutions as well. This demonstrates that the deviations of the parameters calculated by BOXER are typical for the cross sections used. A large number of critical experiments were calculated using the measured material bucklings in order to bring to light possible trends in the calculation of the multiplication factor k/sub eff/. First it came out that the error bounds of B/sub m//sup 2/ evalu-ated in the measurements are often optimistic. Two-dimensional calculations improved the results of the cell calculations. With a mean scattering of 4 to 5 mk in the normal arrays, the multiplication factors calculated by BOXER are satisfactory. However one has to take into account a slight trend of k/sub eff/ to grow with the moderator to fuel ratio and the enrichment. (author)

  10. Recycle of LWR [Light Water Reactor] actinides to an IFR [Integral Fast Reactor

    International Nuclear Information System (INIS)

    Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

    1991-01-01

    A large quantity of actinide elements is present in irradiated Light Water Reactor (LWR) fuel that is stored throughout the world. Because of the high fission-to-capture ratio for the transuranium (TRU) elements with the high-energy neutrons in the metal-fueled Integral Fast Reactor (IFR), that reactor can consume these elements effectively. The stored fuel represents a valuable resource for an expanding application of fast power reactors. In addition, removal of the TRU elements from the spent LWR fuel has the potential for increasing the capacity of a high-level waste facility by reducing the heat loads and increasing the margin of safety in meeting licensing requirements. Argonne National Laboratory (ANL) is developing a pyrochemical process, which is compatible with the IFR fuel cycle, for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. The major objective of the LWR fuel recovery process is high TRU element recovery, with decontamination a secondary issue, because fission product removal is accomplished in the IFR process. The extensive pyrochemical processing studies of the 1960s and 1970s provide a basis for the design of possible processes. Two processes were selected for laboratory-scale investigation. One is based on the Salt Transport Process studied at ANL for mixed-oxide fast reactor fuel, and the other is based on the blanket processing studies done for ANL's second Experimental Breeder Reactor (EBR-2). This paper discusses the two processes and is a status report on the experimental studies. 5 refs., 2 figs., 2 tabs

  11. Quality assurance in the course of fabrication of LWR fuel

    International Nuclear Information System (INIS)

    Dressler, G.; Perry, J.A.

    1982-01-01

    A high quality level of LWR fuel elements can only be assured by a system of Quality Assurance measures purposefully designed, balanced, and appropriately applied. This includes application of and the appropriate balance between both system and product oriented measures. A prerequisite to the establishment of these measures is a precise analysis of the various influences of the individual process steps on the quality characteristics of the starting materials, semi-finished and finished products. In addition, these characteristics require classification criteria relative to their significance. The described classification is used to establish sampling plans and to disposition non-conformances. The EXXON Nuclear Quality Assurance system which is based on these principles is described and illustrated with some examples. (orig.)

  12. Design consideration on severe accident for future LWR

    International Nuclear Information System (INIS)

    Omoto, A.

    1998-01-01

    Utilities' Severe Accident Management strategies, selected based on Individual Plant Examination, are in the process of implementation for each operating plant. Activities for the next generation LWR design are going on by Utilities, NSSS vendors and Research Institutes. The proposed new designs vary from evolutionary design to revolutionary design such as the supercritical LWR. Discussion on the consideration of Severe Accident in the design of next generation LWR is being held to establish the industry's self-regulatory document on containment design and its performance, which ABWR-IER (Improved Evolutionary Reactor) on the part of BWR and Evolutionary APWR and New PWR21 on the part of PWR are expected to comply. Conceptual design study for ABWR-IER will illustrate an example of design approach for the prevention and mitigation of Severe Accident and its impact on capital cost

  13. Assessment of LWR spent fuel disposal options. Volume 3. Study bases and system design considerations (Appendices). Technical report

    Energy Technology Data Exchange (ETDEWEB)

    1979-07-01

    Volume 3 (Appendices) provides a tabulation of the bases and assumptions used in the study as well as preconceptual design description and cost estimates of the facilities and transportation systems necessary to implement the various study cases.

  14. Assessment of LWR spent fuel disposal options. Volume 3. Study bases and system design considerations (Appendices). Technical report

    International Nuclear Information System (INIS)

    1979-07-01

    Volume 3 (Appendices) provides a tabulation of the bases and assumptions used in the study as well as preconceptual design description and cost estimates of the facilities and transportation systems necessary to implement the various study cases

  15. The minimum attention plant inherent safety through LWR simplification

    International Nuclear Information System (INIS)

    Turk, R.S.; Matzie, R.A.

    1987-01-01

    The Minimum Attention Plant (MAP) is a unique small LWR that achieves greater inherent safety, improved operability, and reduced costs through design simplification. The MAP is a self-pressurized, indirect-cycle light water reactor with full natural circulation primary coolant flow and multiple once-through steam generators located within the reactor vessel. A fundamental tenent of the MAP design is its complete reliance on existing LWR technology. This reliance on conventional technology provides an extensive experience base which gives confidence in judging the safety and performance aspects of the design

  16. Recycle of LWR actinides to an IFR

    International Nuclear Information System (INIS)

    Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

    1991-01-01

    Large quantities of actinide elements are present in irradiated light water reactor fuel that is stored throughout the world. Because of the high fission to capture ratio for the transuranium (TRU) elements with the high energy neutrons in metal-fueled integral fast reactors (IFR), that reactor can consume these elements effectively. The stored fuel may represent valuable resource for the expanding application of fast power reactors. In addition, the removal of TRU elements from spent LWR fuel has the potential for increasing the capacity of high level waste facilities by reducing the heat load and may increase the margin of safety in meeting licensing requirement. Argonne National Laboratory is developing a pyrochemical process, which is compatible with the IFR fuel cycle for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. Two pyrochemical processes, that is, salt transport process and blanket processing study, are discussed in this paper. Also the experimental studies are reported. (K.I.)

  17. Validating the BISON fuel performance code to integral LWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

    2016-05-15

    Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

  18. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

  19. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1998-03-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database.

  20. ERDA LWR plant technology program: role of government/industry in improving LWR performance

    International Nuclear Information System (INIS)

    1975-01-01

    Information is presented under the following chapter headings: executive summary; LWR plant outages; LWR plant construction delays and cancellations; programs addressing plant outages, construction delays, and cancellations; need for additional programs to remedy continuing problems; criteria for government role in LWR commercialization; and the proposed government program

  1. Utilities/industries joint study on seismic isolation systems for LWR: Part I. Experimental and analytical studies on seismic isolation systems

    International Nuclear Information System (INIS)

    Kato, Muneaki; Sato, Shoji; Shimomura, Issei

    1989-01-01

    This paper describes a joint study program on seismic isolation systems for light-water reactors (LWRs) performed by ten electric power companies, three manufacturers, and five construction companies. The fundamental response characteristics of base-isolated structures and base-isolation devices are described. Applications of a base-isolation system to LWR buildings are given. Finally, three-dimensional shaking table experiments are described

  2. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-01

    Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

  3. Flexible fuel cycle system for the transition from LWR to FBR

    International Nuclear Information System (INIS)

    Fukasawa, Tetsuo; Yamashita, Junichi; Hoshino, Kuniyoshi; Sasahira, Akira; Inoue, Tadashi; Minato, Kazuo; Sato, Seichi

    2009-01-01

    Japan will deploy commercial fast breeder reactor (FBR) from around 2050 under the suitable conditions for the replacement of light water reactor (LWR) with FBR. The transition scenario from LWR to FBR is investigated in detail and the flexible fuel cycle initiative (FFCI) system has been proposed as a optimum transition system. The FFCI removes ∼95% uranium from LWR spent fuel (SF) in LWR reprocessing and residual material named Recycle Material (RM), which is ∼1/10 volume of original SF and contains ∼50% U, ∼10% Pu and ∼40% other nuclides, is treated in FBR reprocessing to recover Pu and U. If the FBR deployment speed becomes lower, the RM will be stored until the higher speed again. The FFCI has some merits compared with ordinary system that consists of full reprocessing facilities for both LWR and FBR SF during the transition period. The economy is better for FFCI due to the smaller LWR reprocessing facility (no Pu/U recovery and fabrication). The FFCI can supply high Pu concentration RM, which has high proliferation resistance and flexibly respond to FBR introduction rate changes. Volume minimization of LWR SF is possible for FFCI by its conversion to RM. Several features of FFCI were quantitatively evaluated such as Pu mass balance, reprocessing capacities, LWR SF amounts, RM amounts, and proliferation resistance to compare the effectiveness of the FFCI system with other systems. The calculated Pu balance revealed that the FFCI could supply enough but no excess Pu to FBR. These evaluations demonstrated the applicability of FFCI system to the transition period from LWR to FBR cycles. (author)

  4. Materials choices for the advanced LWR steam generators

    International Nuclear Information System (INIS)

    Paine, J.P.N.; Shoemaker, C.E.; McIlree, A.R.

    1987-01-01

    Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In so far as possible, the materials choices for the advanced LWR steam generator avoid the corrosion pitfalls seemingly inherent in the current designs. The EPRI Steam Generator Project staff has recommended materials and design choices for a new steam generator. Based on these recommendations we believe that the advanced LWR steam generators will be much less affected by corrosion and mechanical damage mechanisms than are now experienced

  5. Metholology for the selection of LWR safety R and D projects. Phase I, status report

    International Nuclear Information System (INIS)

    El-Sheikh, K.A.

    1980-03-01

    The objective of the LWR R and D Selection Methodology Program is to develop and demonstrate an R and D selection methodology appropriate for LWR safety technology. This report documents the development work from the program beginning in April, 1979 to the end of Fiscal Year 1979. The scope of work for this period included three tasks; methodology review (Task 1), measures development (Task 2), and methodology development for the first phase of application (Task 3). The accomplishments of these tasks are presented

  6. Recycling U and Pu in LWR

    International Nuclear Information System (INIS)

    Zheng Hualing.

    1986-01-01

    This article, from viewpoints of technical feasibility, safety evaluation and socioeconomic benefit-risk analysis, introduces and comments on history and status of recycling U and Pu in LWR, dealing with reactor, reprocessing, conversion and fuel element fabrication et al. Author has analysed LWR fuel cycle strategies in China and made a proposal

  7. Expert system for estimating LWR plutonium production

    International Nuclear Information System (INIS)

    Sandquist, G.M.

    1988-01-01

    An Artificial Intelligence-Expert System called APES (Analysis of Proliferation by Expert System) has been developed and tested to permit a non proliferation expert to evaluate the capability and capacity of a specified LWR reactor and PUREX reprocessing system for producing and separating plutonium even when system information may be limited and uncertain. APES employs an expert system coded in LISP and based upon an HP-RL (Hewlett Packard-Representational Language) Expert System Shell. The user I/O interface communicates with a blackboard and the knowledge base which contains the quantitative models required to describe the reactor, selected fission product production and radioactive decay processes, Purex reprocessing and ancillary knowledge

  8. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2E, Physical descriptions of LWR nonfuel assembly hardware, Appendix 2F, User's guide to the LWR nonfuel assembly data base

    International Nuclear Information System (INIS)

    1987-12-01

    This appendix includes a two to three page Physical Description report for each Non-fuel Assembly (NFA) Hardware item identified from the current data. Information was obtained via subcontracts with these NFA hardware vendors: Babcock and Wildox, Combustion Engineering and Westinghouse. Data for some NFA hardware are not available. For such hardware, the information shown in this report was obtained from the open literature. Efforts to obtain additional information are continuing. NFA hardware can be grouped into six categories: BWR Channels, Control Elements, Guide Tube Plugs/Orifice Rods, Instrumentation, Neutron Poisons, and Neutron Sources. This appendix lists Physical Description reports alphabetically by vendor within each category. Individual Physical Description reports can be generated interactively through the menu-driven LWR Non-Fuel Assembly Hardware Data Base system. These reports can be viewed on the screen, directed to a printer, or saved in a text file for later use. Special reports and compilations of specific data items can be produced on request

  9. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E., E-mail: luisen.herranz@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Garcia, Monica, E-mail: monica.gmartin@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Morandi, Sonia, E-mail: sonia.morandi@rse-web.it [Nuclear and Industrial Plant Safety Team, Power Generation System Department, RSE, via Rubattino 54, 20134 Milano (Italy)

    2013-12-15

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have

  10. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Morandi, Sonia

    2013-01-01

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have been adopted so that

  11. Outline of Swedish activities on LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M [Studsvik Nuclear, Nykoeping (Sweden); Roennberg, G [OKG AB (Sweden)

    1997-12-01

    The presentation outlines the Swedish activities on LWR fuel and considers the following issues: electricity production; performance of operating nuclear power plants; nuclear fuel cycle and waste management; research and development in nuclear field. 4 refs, 4 tabs.

  12. Implications of plutonium utilization strategies on the transition from a LWR economy to a breeder economy

    International Nuclear Information System (INIS)

    Newman, D.F.; Fleischman, R.M.; White, M.K.

    1977-02-01

    The plutonium interface between the LWR and LMFBR fuel cycles is examined for typical nuclear growth projections both with and without plutonium recycle in LWRs. In order to guarantee a fuel supply for projected LMFBR growth rates, significant multiple Pu recycle in LWRs will not be possible. However, about 78% of the benefit of multiple plutonium recycle between now and the turn of the century is realized by one recycle and then stockpiling spent MOX for the LMFBR. LMFBR reprocessing schecules are estimated based on accumulation of reprocessing load. These schedules are used to estimate the amount of plutonium recovered from LMFBR fuels and determine the residual LWR plutonium required to meet LMFBR demand. The stockpile of LWR produced plutonium in spent MOX is sufficient to fuel the LMFBR until commercial LMFBR reprocessing can be justified. After that time, recycle of plutonium in LWRs will be significantly limited by a continuing LMFBR demand for LWR plutonium due to the projected high LMFBR growth rate. LWR reprocessing requirements are estimated for the assumed condition that LWR plutonium recycle is not approved, but the LMFBR is still pursued as an energy option. The uncertainties presented by this condition are addressed qualitatively. However, in our judgment these uncertainties in the plutonium market would likely delay LMFBR growth to levels significantly below current projections

  13. Safety aspects and operating experience of LWR plants in Japan

    International Nuclear Information System (INIS)

    Aoki, S.; Yoshioka, T.; Toyota, M.; Hinoki, M.

    1977-01-01

    To develop nuclear power generation for the future, it is necessary to put further emphasis on safety assurance and to endeavour to devise measures to improve plant availability, based on the careful analysis of causes that reduce plant availability. The paper discusses the results of studies on the following items from such viewpoints: (1) Safety and operating experience of LWR nuclear power plants in Japan: operating experience with LWRs; improvements in LWR design during the past ten years; analysis of the factors affecting plant availability; (2) Assurance of safety and measures to increase availability: measures for safety and environmental protection; measures to reduce radiation exposure of employees; appropriateness of maintenance and inspection work; measures to increase plant availability; measures to improve reliability of equipment and components; (3) Future technical problems. (author)

  14. LWR aerosol containment experiments (LACE) program and initial test results

    International Nuclear Information System (INIS)

    Muhlestein, L.D.; Hilliard, R.K.; Bloom, G.R.; McCormack, J.D.; Rahn, F.J.

    1985-01-01

    The LWR aerosol containment experiments (LACE) program is described. The LACE program is being performed at the Hanford Engineer Development Laboratory (operated by Westinghouse Hanford Company) and the initial tests are sponsored by EPRI. The objectives of the LACE program are: to demonstrate, at large-scale, inherent radioactive aerosol retention behavior for postulated high consequence LWR accident situations; and to provide a data base to be used for aerosol behavior . Test results from the first phase of the LACE program are presented and discussed. Three large-scale scoping tests, simulating a containment bypass accident sequence, demonstrated the extent of agglomeration and deposition of aerosols occurring in the pipe pathway and vented auxiliary building under realistic accident conditions. Parameters varied during the scoping tests were aerosol type and steam condensation

  15. Tools for LWR spent fuel characterization: Assembly classes and fuel designs

    International Nuclear Information System (INIS)

    Moore, R.S.; Notz, K.J.

    1991-01-01

    The Characteristics Data Base (CDB) is sponsored by the DOE's Office of Civilian Radioactive Waste Management (OCRWM). The CDB provides a single, comprehensive source of data pertaining to radioactive wastes that will or may require geologic disposal, including detailed data describing the physical, quantitative, and radiological characteristics of light-water reactor (LWR) spent fuel. In developing the CDB, tools for the classification of fuel assembly types have been developed. The assembly class scheme is particularly useful for size- and handling-based describes these tools and presents results of their applications in the areas of fuel assembly type identification, characterization of projected discharges, cask accommodation analyses, and defective fuel analyses. Suggestions for additional applications are also made. 7 refs., 1 fig., 2 tabs

  16. Assessment of management alternatives for LWR wastes. Volume 6. Cost determination of the LWR waste management routes (treatment/conditioning/packaging/transport operations)

    International Nuclear Information System (INIS)

    Thiels, G.M.; Kowa, S.

    1993-01-01

    This report deals with the cost determination of a number of schemes for the treatment, conditioning, packaging, interim storage and transport operations of LWR wastes drawn up on the basis of Belgian, French and German practices in this particular area. In addition to the general procedure elaborated for determining, actualizing and scaling of plant and transport costs associated with the various schemes, in-depth calculations of each intermediate management stage are included in this report. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  17. Utility requirements for advanced LWR passive plants

    International Nuclear Information System (INIS)

    Yedidia, J.M.; Sugnet, W.R.

    1992-01-01

    LWR Passive Plants are becoming an increasingly attractive and prominent option for future electric generating capacity for U.S. utilities. Conceptual designs for ALWR Passive Plants are currently being developed by U.S. suppliers. EPRI-sponsored work beginning in 1985 developed preliminary conceptual designs for a passive BWR and PWR. DOE-sponsored work from 1986 to the present in conjunction with further EPRI-sponsored studies has continued this development to the point of mature conceptual designs. The success to date in developing the ALWR Passive Plant concepts has substantially increased utility interest. The EPRI ALWR Program has responded by augmenting its initial scope to develop a Utility Requirements Document for ALWR Passive Plants. These requirements will be largely based on the ALWR Utility Requirements Document for Evolutionary Plants, but with significant changes in areas related to the passive safety functions and system configurations. This work was begun in late 1988, and the thirteen-chapter Passive Plant Utility Requirements Document will be completed in 1990. This paper discusses the progress to date in developing the Passive Plant requirements, reviews the top-level requirements, and discusses key issues related to adaptation of the utility requirements to passive safety functions and system configurations. (orig.)

  18. Economic analyses of LWR fuel cycles

    International Nuclear Information System (INIS)

    Field, F.R.

    1977-05-01

    An economic comparison was made of three options for handling irradiated light-water reactor (LWR) fuel. These options are reprocessing of spent reactor fuel and subsequent recycle of both uranium and plutonium, reprocessing and recycle of uranium only, and direct terminal storage of spent fuel not reprocessed. The comparison was based on a peak-installed nuclear capacity of 507 GWe by CY 2000 and retirement of reactors after 30 years of service. Results of the study indicate that: Through the year 2000, recycle of uranium and plutonium in LWRs saves about $12 billion (FY 1977 dollars) compared with the throwaway cycle, but this amounts to only about 1.3% of the total cost of generating electricity by nuclear power. If deferred costs are included for fuel that has been discharged from reactors but not reprocessed, the economic advantage increases to $17.7 billion. Recycle of uranium only (storage of plutonium) is approximately $7 billion more expensive than the throwaway fuel cycle and is, therefore, not considered an economically viable option. The throwaway fuel cycle ultimately requires >40% more uranium resources (U 3 O 8 ) than does reprocessing spent fuel where both uranium and plutonium are recycled

  19. NUPEC proves reliability of LWR fuel assemblies

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    It is very important in assuring the safety of nuclear reactors to confirm the reliability of fuel assemblies. The test program of the Nuclear Power Engineering Center on the reliability of fuel assemblies has verified the high performance and reliability of Japanese LWR fuels, and confirmed the propriety of their design and fabrication. This claim is based on the data obtained from the fuel assemblies irradiated in commercial reactors. The NUPEC program includes irradiation test which has been conducted for 11 years since fiscal 1976, and the maximum thermal loading test using the out of pile test facilities simulating a real reactor which has been continued since fiscal 1978. The irradiation test on BWR fuel assemblies in No.3 reactor in Fukushima No.1 Nuclear Power Station, Tokyo Electric Power Co., Inc., and on PWR fuel assemblies in No.3 reactor in Mihama Power Station, Kansai Electric Power Co., Inc., and the maximum thermal loading test on BWR and PWR fuel assemblies are reported. The series of postirradiation examination of the fuel assemblies used for commercial reactors was conducted for the first time in Japan, and the highly systematic data on 27 items were obtained. (Kako, I.)

  20. Results of LWR core transient benchmarks

    International Nuclear Information System (INIS)

    Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.

    1993-10-01

    LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs

  1. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  2. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  3. REST based mobile applications

    Science.gov (United States)

    Rambow, Mark; Preuss, Thomas; Berdux, Jörg; Conrad, Marc

    2008-02-01

    Simplicity is the major advantage of REST based webservices. Whereas SOAP is widespread in complex, security sensitive business-to-business aplications, REST is widely used for mashups and end-user centric applicatons. In that context we give an overview of REST and compare it to SOAP. Furthermore we apply the GeoDrawing application as an example for REST based mobile applications and emphasize on pros and cons for the use of REST in mobile application scenarios.

  4. Evaluation of nuclear fuel reprocessing strategies. 2. LWR fuel storage, recycle economics and plutonium logistics

    International Nuclear Information System (INIS)

    Prince, B.E.; Hadley, S.W.

    1983-01-01

    This is the second of a two-part report intended as a critical review of certain issues involved with closing the Light Water Reactor (LWR) fuel cycle and establishing the basis for future transition to commercial breeder applications. The report is divided into four main sections consisting of (1) a review of the status of the LWR spent fuel management and storage problem; (2) an analysis of the economic incentives for instituting reprocessing and recycle in LWRs; (3) an analysis of the time-dependent aspects of plutonium economic value particularly as related to the LWR-breeder transition; and (4) an analysis of the time-dependent aspects of plutonium requirements and supply relative to this transition

  5. Feasibility assessment of the once-through thorium fuel cycle for the PTVM LWR concept

    International Nuclear Information System (INIS)

    Rachamin, R.; Fridman, E.; Galperin, A.

    2015-01-01

    Highlights: • The PTVM LWR is an innovation reactor concept operating in a “breed & burn” mode. • An advanced once-through thorium fuel cycle for the PTVM LWR concept is proposed. • The PTVM LWR concept makes use of a seed-blanket geometry. • A novel fuel management scheme based on two separate fuel flow routes is analyzed. • The analysis indicates a potential for utilizing the fuel in an efficient manner. - Abstract: This paper investigates the feasibility of a once-through thorium fuel cycle for the novel reactor-design concept named the pressure tube light water reactor with variable moderator control (PTVM LWR). The PTVM LWR operates in a “breed & burn” mode, which makes it an attractive system for utilizing thorium fuel in a once-through mode. The “breed & burn” mode can emphasize the in situ generation as well as incineration of 233 U, which are the basic foundations of the once-through thorium fuel cycle. The PTVM LWR concept makes use of a seed–blanket geometry, whereby the core is divided into separated regions of thorium-based fuel channel assemblies (blanket) and low-enriched uranium (LEU) based fuel channel assemblies (seed). A novel fuel in-core management scheme based on two separate fuel flow routes (i.e., seed route and blanket route) is proposed and analyzed. Neutronic performance analysis indicates that the proposed novel fuel in-core management scheme has the potential to utilize both LEU- and thorium-based fuel in an efficient manner. The once-through thorium cycle, presented and discussed in this paper, provide interesting research leads and can serve as a bridge between current LEU-based fuel cycles and a thorium fuel cycle based on recycling of 233 U

  6. Issues in risk analysis of passive LWR designs

    International Nuclear Information System (INIS)

    Youngblood, R.W.; Pratt, W.T.; Amico, P.J.; Gallagher, D.

    1992-01-01

    This paper discusses issues which bear on the question of how safety is to be demonstrated for ''simplified passive'' light water reactor (LWR) designs. First, a very simplified comparison is made between certain systems in today's plants. comparable systems in evolutionary designs, and comparable systems in the simplified passives. in order to introduce the issues. This discussion is not intended to describe the designs comprehensively, but is offered only to show why certain issues seem to be important in these particular designs. Next, an important class of accident sequences is described; finally, based on this discussion, some priorities in risk analysis are presented and discussed

  7. Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1986-01-01

    The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) and ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models

  8. Evaluation of main control room habitability in Japanese LWR (2). Evaluation for applicability of existing atmospheric dispersion models to building wake dispersion by using wind tunnel experiment

    International Nuclear Information System (INIS)

    Fukuda, Ryo; Fujita, Yuko; Yoneda, Jiro; Okabayashi, Kazuki; Tabuse, Shigehiko; Watada, Masayuki

    2009-01-01

    It is necessary to predict the concentration field behind the containment vessel building for the evaluation of main control room habitability in case of the emergency. The concentration field behind the building is very complicated phenomena and the exact prediction of concentration would be very difficult even if philosophical numerical simulation was used. Instead the simple and analytical prediction models (ARCON96, Gifford and Murphy-Campe etc.) have been used for the assessment of main control room habitability. In order to evaluate the previous models, the wind tunnel experiment was carried out. Recent regulatory models of ADMS4 developed by UK-CERC and AERMOD by US-EPA were also compared with this experimental data. Only both the containment vessel and reactor buildings of the typical PWR plant was scaled in 1/200 and the atmospheric stability C-D between C and D of Pasquill-Gifford categories was reproduced as a neutral condition in the wind tunnel experiment. In the wind experiment, the meandering effect for 1 hour was taken into consideration by the so-called overlapping method that a scaled model in the test section of a wind tunnel was rotated. By the rotation of the scaled model, wind directional fluctuations were relatively generated in the test section. The model was rotated at a various speed which was inversely proportional to each frequency of occurrence of a wind direction. Tracer gas was sampled during the rotation of the building model. As a result, we got the 1 hr.-averaged concentration taking a meandering effect into consideration. In this experiment, it is assumed that the frequency distribution of wind direction is Gaussian and horizontal plume width for 1 hr. was expanded to about 1.8 times of plume width based on Pasquill-Gifford chart by 1/5 power law due to the meandering effect. From the experiment, it was found as follows; It seems that meandering effect was not important in the near field behind a building, because strong

  9. Advanced LWR technology for commercial application

    International Nuclear Information System (INIS)

    Redding, J.R.

    1993-01-01

    Advanced Light Water Reactors (ALWRs) are now being deployed and commercialized around the world. In Japan, the Tokyo Electric Power Company (TEPCO) is building the world's first ALWRs, two 1300 MWe Advanced BWRs (ABWRs). In the United States, the Department of Energy, utilities and suppliers are undertaking a cooperative program called First of a Kind Engineering (FOAKE). The purpose of FOAKE is to perform the detailed engineering of ALWRs to that they will be commercially available to U.S. utilities in the mid-1990s. The U.S. industry is in the second year of its strategic plan to have an ALWR in commercial operation by the year 2000. Elsewhere, the Taiwan Power Company has issued a Request for Proposal for two ALWRs so be built at its Lungmen site, with commercial operation of the first unit to be in the year 2000. Korea is formulating plans for an ALWR and other countries, such as Indonesia and Mexico, are looking into the feasibility of building ALWRs

  10. A New Coupled CFD/Neutron Kinetics System for High Fidelity Simulations of LWR Core Phenomena: Proof of Concept

    Directory of Open Access Journals (Sweden)

    Jorge Pérez Mañes

    2014-01-01

    Full Text Available The Institute for Neutron Physics and Reactor Technology (INR at the Karlsruhe Institute of Technology (KIT is investigating the application of the meso- and microscale analysis for the prediction of local safety parameters for light water reactors (LWR. By applying codes like CFD (computational fluid dynamics and SP3 (simplified transport reactor dynamics it is possible to describe the underlying phenomena in a more accurate manner than by the nodal/coarse 1D thermal hydraulic coupled codes. By coupling the transport (SP3 based neutron kinetics (NK code DYN3D with NEPTUNE-CFD, within a parallel MPI-environment, the NHESDYN platform is created. The newly developed system will allow high fidelity simulations of LWR fuel assemblies and cores. In NHESDYN, a heat conduction solver, SYRTHES, is coupled to NEPTUNE-CFD. The driver module of NHESDYN controls the sequence of execution of the solvers as well as the communication between the solvers based on MPI. In this paper, the main features of NHESDYN are discussed and the proof of the concept is done by solving a single pin problem. The prediction capability of NHESDYN is demonstrated by a code-to-code comparison with the DYNSUB code. Finally, the future developments and validation efforts are highlighted.

  11. Advanced hybrid process with solvent extraction and pyro-chemical process of spent fuel reprocessing for LWR to FBR

    International Nuclear Information System (INIS)

    Fujita, Reiko; Mizuguchi, Koji; Fuse, Kouki; Saso, Michitaka; Utsunomiya, Kazuhiro; Arie, Kazuo

    2008-01-01

    Toshiba has been proposing a new fuel cycle concept of a transition from LWR to FBR. The new fuel cycle concept has better economical process of the LWR spent fuel reprocessing than the present Purex Process and the proliferation resistance for FBR cycle of plutonium with minor actinides after 2040. Toshiba has been developing a new Advanced Hybrid Process with Solvent Extraction and Pyrochemical process of spent fuel reprocessing for LWR to FBR. The Advanced Hybrid Process combines the solvent extraction process of the LWR spent fuel in nitric acid with the recovery of high pure uranium for LWR fuel and the pyro-chemical process in molten salts of impure plutonium recovery with minor actinides for metallic FBR fuel, which is the FBR spent fuel recycle system after FBR age based on the electrorefining process in molten salts since 1988. The new Advanced Hybrid Process enables the decrease of the high-level waste and the secondary waste from the spent fuel reprocessing plants. The R and D costs in the new Advanced Hybrid Process might be reduced because of the mutual Pyro-chemical process in molten salts. This paper describes the new fuel cycle concept of a transition from LWR to FBR and the feasibility of the new Advanced Hybrid Process by fundamental experiments. (author)

  12. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    International Nuclear Information System (INIS)

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option

  13. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

  14. Development of training simulator for LWR

    International Nuclear Information System (INIS)

    Sureshbabu, R.M.

    2009-01-01

    A full-scope training simulator was developed for a light water reactor (LWR). This paper describes how the development evolved from a desktop simulator to the full-scope training simulator. It also describes the architecture and features of the simulator including the large number of failures that it simulates. The paper also explains the three-level validation tests that were used to qualify the training simulator. (author)

  15. Modeling the economic consequences of LWR accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Aldrich, D.C.; Rasmussen, N.C.

    1984-01-01

    Models to be used for analyses of economic risks from events which may occur during LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models can be used by both the nuclear power industry and regulatory agencies in cost-benefit analyses for decisionmaking purposes. The newly developed economic consequence models are applied in an example to estimate the economic risks from operation of the Surry Unit 2 plant. The analyses indicate that economic risks from US LWR operation, in contrast to public health risks, are dominated by relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for the Surry site. The implications of these conclusions for nuclear power plant operation and regulation are discussed

  16. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    International Nuclear Information System (INIS)

    Greene, S.R.; Fisher, S.E.; Bevard, B.B.

    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 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

  17. FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    Greene, S.R.; Fisher, S.E.; 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 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

  18. Technical report on LWR design decision methodology. Phase I

    International Nuclear Information System (INIS)

    1980-03-01

    Energy Incorporated (EI) was selected by Sandia Laboratories to develop and test on LWR design decision methodology. Contract Number 42-4229 provided funding for Phase I of this work. This technical report on LWR design decision methodology documents the activities performed under that contract. Phase I was a short-term effort to thoroughly review the curret LWR design decision process to assure complete understanding of current practices and to establish a well defined interface for development of initial quantitative design guidelines

  19. Utilities/industries joint study on seismic isolation systems for LWR: Part II. Observed behaviors of base-isolated general buildings under real earthquakes

    International Nuclear Information System (INIS)

    Matsumura, Takao; Sato, Shoji; Kato, Muneaki

    1989-01-01

    This paper describes the observed behavior of base-isolated buildings under real earthquake conditions. These buildings were constructed by five construction companies participating in the Joint Study on Seismic Isolation Systems for lightwater reactors. All the buildings are medium- or low-height buildings of reinforced-concrete structures with combinations of laminated rubber bearing or sliding bearings and various damping devices

  20. LWR risk management by safety R and D

    International Nuclear Information System (INIS)

    El-Sheikh, K.A.; Damon, D.R.; Temme, M.I.

    1982-01-01

    This paper presents a methodology which has been developed for selecting LWR safety RandD projects. The methodology provides ranking of the RandD projects and the RandD budget allocation which minimizes public risk. The methodology contains procedures to identify institutional, organizational, legal, and contractual factors which affect the probabilities of success and use of RandD projects so that these factors can be evaluated and possibly managed.The methodology also contains a nonlinear optimization code to provide the optimum selection of RandD projects and evaluate the sensitivity of this selection to uncertainity in the input data. Application of the methodology to a test case has shown that: 1) commonly used schemes for ranking RandD projects do not necessarily lead to the optimum selection, and 2) the optimum selection is not necessarily strongly sensitive to uncertainty in the input data

  1. Fracture probability evaluation of a LWR pressure vessel

    International Nuclear Information System (INIS)

    Grandemange, J.; Pellissier-Tanon, A.; Quero, J.; Carnino, A.; Dufresne, J.

    1978-01-01

    Fracture probability evaluation, of a LWR pressure vessel have been performed in the past, using statistical data from conventional plant. A more accurate evaluation has been requested in 1976 from the SCSIN to the CEA. With this object, a joint collaboration agreement has been signed between CEA, EURATOM/ISPRA and FRAMATOME. The whole program proceeding from this agreement is managed by a joint board including the three partners. The basic objective of this program is to develop a method which integrates, or makes it possible to integrate at a later stage, the greatest number of significant parameters. Also, in order to prepare the practical applications, a special effort is being made to collect the data corresponding to these parameters. Parallel basic research program have been launched in order to clarify our knowledge on some important parts of the main factors contributing to the evaluation. The results of this research will be progressively introduced into the method or will help checking its validity

  2. Spent LWR fuel leach tests: Waste Isolation Safety Assessment program

    International Nuclear Information System (INIS)

    Katayama, Y.B.

    1979-04-01

    Spent light-water-reactor (LWR) fuels with burnups of 54.5, 28 and 9 MWd/kgU were leach-tested in deionized water at 25 0 C. Fuel burnup has no apparent effect on the calculated leach rates based upon the behavior of 137 Cs and 239+240 Pu. A leach test of 54.5 MWd/kgU spent fuel in synthetic sea brine showed that the cesium-based leach rate is lower in sea brine than in deionized water. A rise in the leach rate was observed after approximately 600 d of cumulative leaching. During the rise, the leach rate for all the measured radionuclides become nearly equal. Evidence suggests that exposure of new surfaces to the leachant may cause the increase. As a result, experimental work to study leaching mechanisms of spent fuel has been initiated. 22 figures

  3. Automatic test equipment for C and I of compact LWR

    International Nuclear Information System (INIS)

    Mayya, Anuradha; Marathe, P.P.; Madala, Kalyan C.

    2014-01-01

    The C and I of compact LWR consist of a wide variety of electronic modules. Testing of these modules manually was found to be very cumbersome. To ease the testing of these modules, Automatic Test Equipments (ATE) were developed jointly by BARC and ECIL. This paper describes the design of two ATEs for testing 69 types of modules. A power supply ATE was developed for 43 types of power supply modules of type AC-AC, AC-DC, DC-DC and signal conditioning modules. A VME ATE was developed to test 26 types of VME bus based and other microcontroller based non-bussed modules. These ATEs are used for the automated black box testing of modules by feeding power and control inputs and checking the outputs without operator intervention. This paper describes the important considerations in design and the major design challenges. (author)

  4. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    International Nuclear Information System (INIS)

    Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki

    2012-01-01

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  5. Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Permana, Sidik; Suzuki, Mitsutoshi; Su' ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2012-06-06

    Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

  6. WAP - based telemedicine applications

    International Nuclear Information System (INIS)

    Hung, K.; Zhang, Y.T.

    2001-01-01

    Telemedicine refers to the utilization of telecommunication technology for medical diagnosis, treatment, and patient care. Its aim is to provide expert-based health care to remote sites through telecommunication and information technologies. The significant advances in technologies have enabled the introduction of a broad range of telemedicine applications, which are supported by computer networks, wireless communication, and information superhighway. For example, some hospitals are using tele-radiology for remote consultation. Such a system includes medical imaging devices networked with computers and databases. Another growing area is patient monitoring, in which sensors are used to acquire biomedical signals, such as electrocardiogram (ECG), blood pressure, and body temperature, from a remote patient, who could be in bed or moving freely. The signals are then relayed to remote systems for viewing and analysis. Telemedicine can be divided into two basic modes of operations: real-time mode, in which the patient data can be accessed remotely in real-time, and store-and-forward mode, in which the acquired data does not have to be accessed immediately. In the recent years, many parties have demonstrated various telemedicine applications based on the Internet and cellular phone as these two fields have been developing rapidly. A current, recognizable trend in telecommunication is the convergence of wireless communication and computer network technologies. This has been reflected in recently developed telemedicine systems. For example, in 1998 J. Reponen, et al. have demonstrated transmission and display of computerized tomography (CT) examinations using a remote portable computer wirelessly connected to a computer network through TCP/IP on a GSM cellular phone. Two years later, they carried out the same tests with a GSM-based wireless personal digital assistant (PDA). The WAP (Wireless Application Protocol) Forum was founded in 1997 to create a global protocol

  7. LWR development in the USA

    International Nuclear Information System (INIS)

    Taylor, J.J.; Stahlkopf, K.E.; Noble, D.M.; Dau, G.J.

    1987-01-01

    The U.S. Nuclear Power Industry is developing improved reactor systems to be repared for the expected need of new base load electric generating capacity in the 1990s. The approach being taken is to build upon the large base of existing light water reactor technology, making evolutionary improvements, and finding innovative ways to simplify, shorten construction time, and reduce cost. The U.S. reactor manufacturers, working in collaboration with Japanese utilities, are developing 1300 MWe improved designs. This paper reviews the EPRI ALWR Program which is coordinated with these efforts by the reactor manufactureres. Emphasis is given to that portion of the EPRI Program dealing with conceptual design of smaller rated Advanced Simplified LWRs using passive system design to accomplish major simplification. (author)

  8. Analysis of alternative light water reactor (LWR) fuel cycles

    International Nuclear Information System (INIS)

    Heeb, C.M.; Aaberg, R.L.; Boegel, A.J.; Jenquin, U.P.; Kottwitz, D.A.; Lewallen, M.A.; Merrill, E.T.; Nolan, A.M.

    1979-12-01

    Nine alternative LWR fuel cycles are analyzed in terms of the isotopic content of the fuel material, the relative amounts of primary and recycled material, the uranium and thorium requirements, the fuel cycle costs and the fraction of energy which must be generated at secured sites. The fuel materials include low-enriched uranium (LEU), plutonium-uranium (MOX), highly-enriched uranium-thorium (HEU-Th), denatured uranium-thorium (DU-Th) and plutonium-thorium (Pu-Th). The analysis is based on tracing the material requirements of a generic pressurized water reactor (PWR) for a 30-year period at constant annual energy output. During this time period all the created fissile material is recycled unless its reactivity worth is less than 0.2% uranium enrichment plant tails

  9. RETRANS, Reactivity Transients in LWR

    International Nuclear Information System (INIS)

    Kamelander, G.

    1989-01-01

    1 - Description of program or function: RETRANS is appropriate to calculate power excursions in light water reactors initiated by reactivity insertions due to withdrawal of control elements. As in the code TWIGL, the neutron physics model is based on the time-dependent two-group neutron diffusion equations. The equation of state of the coolant is approximated by a table built into the code. RETRANS solves the heat conduction equation and calculates the heat transfer coefficient for representative fuel rods at each time-step. 2 - Method of solution: The time-dependent neutron diffusion equations are modified by an exponential transformation and solved by means of a finite difference method. There is an option accelerating the inner iterations of the difference scheme by a coarse-mesh re-balancing method. The heat balance equations of the thermo- hydraulic model are discretized and converted into a tri-diagonal system of linear equations which is solved recursively. 3 - Restrictions on the complexity of the problem: r-z-geometry, one- phase-flow

  10. Draft report: a selection methodology for LWR safety R and D programs and proposals

    Energy Technology Data Exchange (ETDEWEB)

    Husseiny, A. A.; Ritzman, R. L.

    1980-03-01

    The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application.

  11. Draft report: a selection methodology for LWR safety R and D programs and proposals

    International Nuclear Information System (INIS)

    Husseiny, A.A.; Ritzman, R.L.

    1980-03-01

    The results of work done to develop a methodology for selecting LWR safety R and D programs and proposals is described. A critical survey of relevant decision analysis methods is provided including the specifics of multiattribute utility theory. This latter method forms the basis of the developed selection methodology. Details of the methodology and its use are provided along with a sample illustration of its application

  12. New development in nondestructive measurement and verification of irradiated LWR fuels

    International Nuclear Information System (INIS)

    Lee, D.M.; Phillips, J.R.; Halbig, J.K.; Hsue, S.T.; Lindquist, L.O.; Ortega, E.M.; Caine, J.C.; Swansen, J.; Kaieda, K.; Dermendjiev, E.

    1979-01-01

    Nondestructive techniques for characterizing irradiated LWR fuel assemblies are discussed. This includes detection systems that measure the axial activity profile, neutron yield and gamma yield. A multi-element profile monitor has been developed that offers a significant improvement in speed and complexity over existing mechanical scanning systems. New portable detectors and electronics, applicable to safeguard inspection, are presented and results of gamma-ray and neutron measurements at commercial reactor facilities are given

  13. Performance and reliability of LWR fuel

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vandenberg, C.

    1977-01-01

    The main requirements for fuel reloads are: good reliability, minimum fuel cycle costs and flexibility of operation. Fulfilling these goals requires a background of experience. The approach to the acquisition of this experience in the particular case of BN has included over the last 15 years a proper development and cross-checking of the design methods and criteria, a continuous updating of the drawings and specifications and the qualification of adequate fabrication plants. This approach can best be outlined on the basis of the gradual implementation of the modern features of the LWR fuel. The first fuel clad with stainless steel was loaded in the BR 3 (11 MWe) in 1969 and later on (since 1974) in the SENA plant (310 MWe). Similarly, Zircaloy 4 cladding was first introduced in a reactor reload in 1969 as autoclaved cladding and later on (in 1971) the autoclaving was suppressed for the further reloads. Zircaloy 2 was loaded in DODEWAARD (51.5 MWe) in 1970. The first demonstration assembly in a PWR was a Pu-island assembly loaded in the BR 3 in 1963. It was followed by an all-Pu assembly in the same reactor in 1965 and by the loading of Pu fuels in four prototype assemblies in GARIGLIANO (160 MWe) in 1968. A full reload incorporating Pu fuel has been experienced by the supply of fuel for GARIGLIANO (BOL: 1975) and for BR 3 (BOL: 1972 and 1976). While in the early sixties the brazed design was still being utilized, the first assembly incorporating grids with springs was introduced in BR 3 in 1963. The first Inconel grids were loaded in the same reactor in 1969 and the first Zircaloy grids in 1972 (the first Zr grid has been loaded in a BWR in 1973). The experience covered successively the shrouded design (BOL: 1963), the shroudless design (BOL: 1969), a BWR assembly (BOL: 1971), a typical RCC assembly first with large diameter fuel rods (1972) and later on with small diameter fuel rods (1974). The experience on the reactivity control covered successively diluted

  14. Ice Thermal Storage Systems for LWR Supplemental Cooling and Peak Power Shifting

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Phil Sharpe; Blaise Hamanaka; Wei Yan; WoonSeong Jeong

    2010-06-01

    benefits and avoid forced derating and shutdown during extremely hot weather. For the new plants using dry cooling towers, adding the ice thermal storage systems can effectively reduce the efficiency loss and water consumption during hot weather so that new LWRs could be considered in regions without enough cooling water. \\ This paper presents the feasibility study of using ice thermal storage systems for LWR supplemental cooling and peak power shifting. LWR cooling issues and ITS application status will be reviewed. Two ITS application case studies will be presented and compared with alternative options: one for once-through cooling without enough cooling for short time, and the other with dry cooling. Because capital cost, especially the ice storage structure/building cost, is the major cost for ITS, two different cost estimation models are developed: one based on scaling method, and the other based on a preliminary design using Building Information Modeling (BIM), an emerging technology in Architecture/Engineering/Construction, which enables design options, performance analysis and cost estimating in the early design stage.

  15. LIFE vs. LWR: End of the Fuel Cycle

    International Nuclear Information System (INIS)

    Farmer, J.C.; Blink, J.A.; Shaw, H.F.

    2008-01-01

    The worldwide energy consumption in 2003 was 421 quadrillion Btu (Quads), and included 162 quads for oil, 99 quads for natural gas, 100 quads for coal, 27 quads for nuclear energy, and 33 quads for renewable sources. The projected worldwide energy consumption for 2030 is 722 quads, corresponding to an increase of 71% over the consumption in 2003. The projected consumption for 2030 includes 239 quads for oil, 190 quads for natural gas, 196 quads for coal, 35 quads for nuclear energy, and 62 quads for renewable sources (International Energy Outlook, DOE/EIA-0484, Table D1 (2006) p. 133]. The current fleet of light water reactors (LRWs) provides about 20% of current U.S. electricity, and about 16% of current world electricity. The demand for electricity is expected to grow steeply in this century, as the developing world increases its standard of living. With the increasing price for oil and gasoline within the United States, as well as fear that our CO2 production may be driving intolerable global warming, there is growing pressure to move away from oil, natural gas, and coal towards nuclear energy. Although there is a clear need for nuclear energy, issues facing waste disposal have not been adequately dealt with, either domestically or internationally. Better technological approaches, with better public acceptance, are needed. Nuclear power has been criticized on both safety and waste disposal bases. The safety issues are based on the potential for plant damage and environmental effects due to either nuclear criticality excursions or loss of cooling. Redundant safety systems are used to reduce the probability and consequences of these risks for LWRs. LIFE engines are inherently subcritical, reducing the need for systems to control the fission reactivity. LIFE engines also have a fuel type that tolerates much higher temperatures than LWR fuel, and has two safety systems to remove decay heat in the event of loss of coolant or loss of coolant flow. These features of

  16. Core design of super LWR with double tube water rods

    International Nuclear Information System (INIS)

    Wu, Jianhui; Oka, Yoshiaki

    2014-01-01

    Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

  17. Coupled Tort-TD/CTF Capability for high-fidelity LWR core calculations - 321

    International Nuclear Information System (INIS)

    Christienne, M.; Avramova, M.; Perin, Y.; Seubert, A.

    2010-01-01

    This paper describes the developed coupling scheme between TORT-TD and CTF. TORT-TD is a time-dependent 3D discrete ordinates neutron transport code. TORT-TD is utilized for high-fidelity reactor core neutronics calculations while CTF is providing the thermal-hydraulics feedback information. CTF is an improved version of the advanced thermal-hydraulic sub-channel code COBRA-TF, which is widely used for best-estimate evaluations of LWR safety margins. CTF is a transient code based on a separated flow representation of the two-phase flow. The coupled code TORT-TD/CTF allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. Steady-state and transient test cases, based on the OECD/NRC PWR MOX/UO 2 Core Transient Benchmark, have been calculated. The steady state cases are based on a quarter core model while the transient test case models a control rod ejection transient in a small PWR mini-core fuel assembly arrangement. The obtained results with TORT-TD/CTF are verified by a code-to-code comparison with the previously developed NEM/CTF and TORT-TD/ATHLET coupled code systems. The performed comparative analysis indicates the applicability and high-fidelity potential of the TORT-TD/CTF coupling. (authors)

  18. A study on gap heat transfer of LWR fuel rods under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Fujishiro, Toshio

    1984-03-01

    Gap heat transfer between fuel pellet and cladding have a large influence on the LWR fuel behaviors under reactivity initiated accident (RIA) conditions. The objective of the present study is to investigate the effects of gap heat transfer on RIA fuel behaviors based on the results of the gap-gas parameter tests in NSRR and on their analysis with NSR-77 code. Through this study, transient variations of gap heat transfer, the effects of the gap heat transfer on fuel thermal behaviors and on fuel failure, effects of pellet-cladding sticking by eutectic formation, and the effects of cladding collapse under high external pressure have been clearified. The studies have also been performed on the applicability and its limit of modified Ross and Stoute equation which is extensively utilized to evaluate the gap heat transfer coefficient in the present fuel behavior codes. The method to evaluate the gap conductance to the conditions beyond the applicability limit of the Ross and Stoute equation has also been proposed. (author)

  19. LWR Spent Fuel Management for the Smooth Deployment of FBR

    International Nuclear Information System (INIS)

    Fukasawa, T.; Yamashita, J.; Hoshino, K.; Sasahira, A.; Inoue, T.; Minato, K.; Sato, S.

    2015-01-01

    Fast breeder reactors (FBR) and FBR fuel cycle are indispensable to prevent the global warming and to secure the long-term energy supply. Commercial FBR expects to be deployed from around 2050 until around 2110 in Japan by the replacement of light water reactors (LWR) after their 60 years life. The FBR deployment needs Pu (MOX) from the LWR-spent fuel (SF) reprocessing. As Japan can posses little excess Pu, its balance control is necessary between LWR-SF management (reprocessing) and FBR deployment. The fuel cycle systems were investigated for the smooth FBR deployment and the effectiveness of proposed flexible system was clarified in this work. (author)

  20. Safety research for LWR type reactors

    International Nuclear Information System (INIS)

    1989-07-01

    The current R and D activities are to be seen in connection with the LWR risk assessment studies. Two trends are emerging, of which the one concentrates more on BWR-specific problems, and the other on the efficiency or safety-related assessment of accident management activities. This annual report of 1988 reviews the progress of work done by the institutes and departments of the Karlsruhe Nuclear Research Center, (KfK), or on behalf of KfK by external institutions, in the field of safety research. The papers of this report present the state of work at the end of the year 1988. They are written in German, with an abstract in English. (orig./HP) [de

  1. LWR nuclear power plant component failures

    International Nuclear Information System (INIS)

    Schmidt, W.H.

    1980-10-01

    An analysis of the most significant light water reactor (LWR) nuclear power plant component failures, from information in the computerized Nuclear Safety Information Center (NSIC) data bank, shows that for both pressurized water reactor (PWR) and boiling water reactor (BWR) plants the component category most responsible for reactor shutdowns is valves. Next in importance for PWR shutdowns is steam generators followed by seals of all kinds. For BWR plants, seals, and pipes and pipe fittings are the second and third most important component failure categories which lead to reactor shutdown. The data are for records extending from early 1972 through September 1978. A list of the most significant component categories and a breakdown of the number of component citations for both PWR and BWR reactor types are presented

  2. Problems associated with domestic LWR technology development

    International Nuclear Information System (INIS)

    Watamori, Tikara

    1975-01-01

    To cope with the future energy problem in Japan, the enhancement of her own technology is continuing in the nuclear power field. Developments in the past, current state, and problems for the future are described regarding LWR power plants. The technology introduced from overseas countries cannot be used as it is. The domestic technology thus consists of the conversion of nuclear power technology so as to meet Japan's own condition and the domestic manufacture of machinery. In the former category, there are the aspects of aseismatic design, waste disposal, software, etc. In the latter, there are the productions of reactor vessels, steam generators, large valves, piping, etc. As the problems for the future, there are reliability and safety and the associated standardization. (Mori, K.)

  3. A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

    International Nuclear Information System (INIS)

    Kim, Chang Hyun

    1997-02-01

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

  4. Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, S. J. [ed.

    1980-04-01

    The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics listed are based upon the experience of ODAI and also based upon a sampling of over 100 engineers/scientists in the LWR industry. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation.

  5. Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

    International Nuclear Information System (INIS)

    Brown, S.J.

    1980-04-01

    The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics listed are based upon the experience of ODAI and also based upon a sampling of over 100 engineers/scientists in the LWR industry. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation

  6. LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3

    International Nuclear Information System (INIS)

    Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.

    1986-09-01

    The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs

  7. Analysis of fission gas release in LWR fuel using the BISON code

    Energy Technology Data Exchange (ETDEWEB)

    G. Pastore; J.D. Hales; S.R. Novascone; D.M. Perez; B.W. Spencer; R.L. Williamson

    2013-09-01

    Recent advances in the development of the finite-element based, multidimensional fuel performance code BISON of Idaho National Laboratory are presented. Specifically, the development, implementation and testing of a new model for the analysis of fission gas behavior in LWR-UO2 fuel during irradiation are summarized. While retaining a physics-based description of the relevant mechanisms, the model is characterized by a level of complexity suitable for application to engineering-scale nuclear fuel analysis and consistent with the uncertainties pertaining to some parameters. The treatment includes the fundamental features of fission gas behavior, among which are gas diffusion and precipitation in fuel grains, growth and coalescence of gas bubbles at grain faces, grain growth and grain boundary sweeping effects, thermal, athermal, and transient gas release. The BISON code incorporating the new model is applied to the simulation of irradiation experiments from the OECD/NEA International Fuel Performance Experiments database, also included in the IAEA coordinated research projects FUMEX-II and FUMEX-III. The comparison of the results with the available experimental data at moderate burn-up is presented, pointing out an encouraging predictive accuracy, without any fitting applied to the model parameters.

  8. Perspectives on the economic risks of LWR accidents

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Burke, R.P.

    1986-01-01

    Models which can be used for the analysis of the economic risks from events which may occur during LWR operation have been developed. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The economic consequence models have been applied in studies of the economic risks from the operation of US LWR plants. The results of the analyses provide some important perspectives regarding the economic risks of LWR accidents. The analyses indicate that economic risks, in contrast to public health risks, are dominated by the onsite costs of relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for a typical US plant

  9. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1983-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  10. Status of the CONTAIN computer code for LWR containment analysis

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.

    1982-01-01

    The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment

  11. Challenges in coupled thermal-hydraulics and neutronics simulations for LWR safety analysis

    International Nuclear Information System (INIS)

    Ivanov, Kostadin; Avramova, Maria

    2007-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal-hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal-hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical

  12. Location Based Application Availability

    Science.gov (United States)

    Naeem Akram, Raja; Markantonakis, Konstantinos; Mayes, Keith

    Smart cards are being integrated into a diverse range of industries: ranging from banking, telecom, transport, home/office access control to health and E-passport. Traditionally, cardholders are required to carry a smart card for each application. However, recent developments in the Near Field Communication (NFC) have renewed the interest in multiple applications for different services on a single device. This paper builds onto the NFC initiative and avoids the smart card ownership issues that hinder the adoption of such devices. The proposal integrates the Global Positioning System with the NFC in mobile phones to provide a ubiquitously and flexible service access model.

  13. Rate Theory Modeling and Simulation of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Gamble, Kyle [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Mei, Zhi-Gang [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-08-29

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well understood. In this report, rate theory model was developed based on existing experimental data and density functional theory (DFT) calculations so as to predict the fission gas behavior in U3Si2 at LWR conditions. The fission gas behavior of U3Si2 can be divided into three temperature regimes. During steady-state operation, the majority of the fission gas stays in intragranular bubbles, whereas the dominance of intergranular bubbles and fission gas release only occurs beyond 1000 K. The steady-state rate theory model was also used as reference to establish a gaseous swelling correlation of U3Si2 for the BISON code. Meanwhile, the overpressurized bubble model was also developed so that the fission gas behavior at LOCA can be simulated. LOCA simulation showed that intragranular bubbles are still dominant after a 70 second LOCA, resulting in a controllable gaseous swelling. The fission gas behavior of U3Si2 at LWR conditions is benign according to the rate theory prediction at both steady-state and LOCA conditions, which provides important references to the qualification of U3Si2 as a LWR fuel material with excellent fuel performance and enhanced accident tolerance.

  14. Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

    International Nuclear Information System (INIS)

    Plaschy, M.; Murphy, M.; Jatuff, F.; Seiler, R.; Chawla, R.

    2006-01-01

    The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of the PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k eff (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)

  15. Evaluation of the 252Cf-source-driven neutron noise analysis method for measuring the subcriticality of LWR fuel storage casks

    International Nuclear Information System (INIS)

    Mihalczo, J.T.

    1987-01-01

    The 252 Cf-source-driven neutron noise analysis method was evaluated to determine if it could be used to measure the subcriticality of storage casks of burnt LWR fuel submerged in fuel storage pools, fully loaded and as they are being loaded. The motivation for this evaluation was that measurements of k/sub eff/ would provide the parameter most directly related to the criticality safety of storage cask configurations of LWR fuel and could allow proper credit for fuel burnup without reliance on calculations. This in turn could lead to more cost-effective cask designs. Evaluation of the method for this application was based on (1) experiments already completed at a critical experiments facility using arrays of PWR fuel pins typical of the size of storage cask configurations, (2) the existence of neutron detectors that can function in shipping cask environments, and (3) the ability to construct ionization chambers containing 252 Cf of adequate intensity for these measurements. These three considerations are discussed

  16. Effects of LWR coolant environments on fatigue lives of austenitic stainless steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Gavenda, D.J.

    1997-01-01

    The ASME Boiler and Pressure Vessel Code fatigue design curves for structural materials do not explicitly address the effects of reactor coolant environments on fatigue life. Recent test data indicate a significant decrease in fatigue life of pressure vessel and piping materials in light water reactor (LWR) environments. Fatigue tests have been conducted on Types 304 and 316NG stainless steel in air and LWR environments to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on fatigue lives of these steels. The results confirm the significant decrease in fatigue life in water. The environmentally assisted decrease in fatigue life depends both on strain rate and DO content in water. A decrease in strain rate from 0.4 to 0.004%/s decreases fatigue life by a factor of ∼ 8. However, unlike carbon and low-alloy steels, environmental effects are more pronounced in low-DO than in high-DO water. At ∼ 0.004%/s strain rate, reduction in fatigue life in water containing <10 ppb D is greater by a factor of ∼ 2 than in water containing ≥ 200 ppb DO. Experimental results have been compared with estimates of fatigue life based on the statistical model. The formation and growth of fatigue cracks in austenitic stainless steels in air and LWR environments are discussed

  17. Criticality impacts on LWR fuel storage efficiency

    International Nuclear Information System (INIS)

    Napolitano, D.

    1992-01-01

    This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

  18. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R and D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. (author)

  19. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  20. Preliminary concepts for detecting diversion of LWR spent fuel

    International Nuclear Information System (INIS)

    Sellers, T.A.

    Sandia Laboratories, under the sponsorship of the Department of Energy, Office of Safeguards and Security, has been developing conceptual designs of advanced systems to rapidly detect diversion of LWR spent fuel. Three detection options have been identified and compared on the basis of timeliness of detection and cost. Option 1 is based upon inspectors visiting each facility on a periodic basis to obtain and review data acquired by surveillance instruments and to verify the inventory. Option 2 is based upon continuous inspector presence, aided by surveillance instruments. Option 3 is based upon the collection of data from surveillance instruments with periodic readout either at the facility or at a remote central monitoring and display module and occasional inspection. Surveillance instruments are included in each option to assure a sufficiently high probability of detection. An analysis technique with an example logic tree that was used to identify performance requirements is described. A conceptual design has been developed for Option 3 and the essential hardware elements are not being developed. These elements include radiation, crane and pool acoustic sensors, a Data Collection Module, a Local Collection Module, a Local Display Module and a Central Monitoring and Display Module. A demonstration, in operating facilities, of the overall system concept is planned for the March to June 1979 time frame

  1. Evaluation of inorganic sorbent treatment for LWR coolant process streams

    International Nuclear Information System (INIS)

    Roddy, J.W.

    1984-03-01

    This report presents results of a survey of the literature and of experience at selected nuclear installations to provide information on the feasibility of replacing organic ion exchangers with inorganic sorbents at light-water-cooled nuclear power plants. Radioactive contents of the various streams in boiling water reactors and pressurized water reactors were examined. In addition, the methods and performances of current methods used for controlling water quality at these plants were evaluated. The study also includes a brief review of the physical and chemical properties of selected inorganic sorbents. Some attributes of inorganic sorbents would be useful in processing light water reactor (LWR) streams. The inorganic resins are highly resistant to damage from ionizing radiation, and their exchange capacities are generally equivalent to those of organic ion exchangers. However, they are more limited in application, and there are problems with physical integrity, especially in acidic solutions. Research is also needed in the areas of selectivity and anion removal before inorganic sorbents can be considered as replacements for the synthetic organic resins presently used in LWRs. 11 figures, 14 tables

  2. Research on ultrasonic flow detection techniques for LWR facilities

    International Nuclear Information System (INIS)

    Kimura, Katsumi; Fukuhara, Hiroaki; Hoshimoto, Kenichi; Matsumoto, Shojiro; Yamawaki, Hisashi; Ito, Hideyuki; Uetake, Ichizo

    1986-01-01

    Aiming at establishing the techniques for inspecting the inside of LWR pressure vessels by ultrasonic flaw detection from the outside of the vessels, the development of a probe suitable to the flaw detection in the thick steel plates with stainless steel overlay and the method of its driving, the examination of the ultrasonic characteristics of austenitic stainless steel welded metal used for overlay, and the improvement of the detectability of defects and the accuracy of measuring dimensions by the application of signal processing techniques to ultrasonic flaw detection were attempted. In order to cope with the impedance lowering accompanying the increase of oscillator size, the oscillator was divided into the rings with equal area, and the driving and signal receiving were carried out individually, in this way, the good results were obtained by summing the signals. It was theoretically proved that it is rational to use longitudinal waves for the flaw detection in overlay. It was found that by displaying the results of flaw detection as pictures using a microcomputer, the capability of defect detection was increased. Also by the signal processing combining Fourier transformation and filtering, noise removal and the heightening of the accuracy of measuring dimensions were able to be attained. (Kako, I.)

  3. The scale analysis sequence for LWR fuel depletion

    International Nuclear Information System (INIS)

    Hermann, O.W.; Parks, C.V.

    1991-01-01

    The SCALE (Standardized Computer Analyses for Licensing Evaluation) code system is used extensively to perform away-from-reactor safety analysis (particularly criticality safety, shielding, heat transfer analyses) for spent light water reactor (LWR) fuel. Spent fuel characteristics such as radiation sources, heat generation sources, and isotopic concentrations can be computed within SCALE using the SAS2 control module. A significantly enhanced version of the SAS2 control module, which is denoted as SAS2H, has been made available with the release of SCALE-4. For each time-dependent fuel composition, SAS2H performs one-dimensional (1-D) neutron transport analyses (via XSDRNPM-S) of the reactor fuel assembly using a two-part procedure with two separate unit-cell-lattice models. The cross sections derived from a transport analysis at each time step are used in a point-depletion computation (via ORIGEN-S) that produces the burnup-dependent fuel composition to be used in the next spectral calculation. A final ORIGEN-S case is used to perform the complete depletion/decay analysis using the burnup-dependent cross sections. The techniques used by SAS2H and two recent applications of the code are reviewed in this paper. 17 refs., 5 figs., 5 tabs

  4. Discussion of OECD LWR Uncertainty Analysis in Modelling Benchmark

    International Nuclear Information System (INIS)

    Ivanov, K.; Avramova, M.; Royer, E.; Gillford, J.

    2013-01-01

    The demand for best estimate calculations in nuclear reactor design and safety evaluations has increased in recent years. Uncertainty quantification has been highlighted as part of the best estimate calculations. The modelling aspects of uncertainty and sensitivity analysis are to be further developed and validated on scientific grounds in support of their performance and application to multi-physics reactor simulations. The Organization for Economic Co-operation and Development (OECD) / Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC) has endorsed the creation of an Expert Group on Uncertainty Analysis in Modelling (EGUAM). Within the framework of activities of EGUAM/NSC the OECD/NEA initiated the Benchmark for Uncertainty Analysis in Modelling for Design, Operation, and Safety Analysis of Light Water Reactor (OECD LWR UAM benchmark). The general objective of the benchmark is to propagate the predictive uncertainties of code results through complex coupled multi-physics and multi-scale simulations. The benchmark is divided into three phases with Phase I highlighting the uncertainty propagation in stand-alone neutronics calculations, while Phase II and III are focused on uncertainty analysis of reactor core and system respectively. This paper discusses the progress made in Phase I calculations, the Specifications for Phase II and the incoming challenges in defining Phase 3 exercises. The challenges of applying uncertainty quantification to complex code systems, in particular the time-dependent coupled physics models are the large computational burden and the utilization of non-linear models (expected due to the physics coupling). (authors)

  5. Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

    Science.gov (United States)

    Snead, L. L.; Terrani, K. A.; Katoh, Y.; Silva, C.; Leonard, K. J.; Perez-Bergquist, A. G.

    2014-05-01

    This paper addresses certain key feasibility issues facing the application of SiC-matrix microencapsulated fuels for light water reactor application. Issues addressed are the irradiation stability of the SiC-based nano-powder ceramic matrix under LWR-relevant irradiation conditions, the presence or extent of reaction of the SiC matrix with zirconium-based cladding, the stability of the inner and outer pyrolytic graphite layers of the TRISO coating system at this uncharacteristically low irradiation temperature, and the state of the particle-matrix interface following irradiation which could possibly affect thermal transport. In the process of determining these feasibility issues microstructural evolution and change in dimension and thermal conductivity was studied. As a general finding the SiC matrix was found to be quite stable with behavior similar to that of CVD SiC. In magnitude the irradiation-induced swelling of the matrix material was slightly higher and irradiation-degraded thermal conductivity was slightly lower as compared to CVD SiC. No significant reaction of this SiC-based nano-powder ceramic matrix material with Zircaloy was observed. Irradiation of the sample in the 320-360 °C range to a maximum dose of 7.7 × 1025 n/m2 (E > 0.1 MeV) did not have significant negative impact on the constituent layers of the TRISO coating system. At the highest dose studied, layer structure and interface integrity remained essentially unchanged with good apparent thermal transport through the microsphere to the surrounding matrix.

  6. Stability of SiC-matrix microencapsulated fuel constituents at relevant LWR conditions

    International Nuclear Information System (INIS)

    Snead, L.L.; Terrani, K.A.; Katoh, Y.; Silva, C.; Leonard, K.J.; Perez-Bergquist, A.G.

    2014-01-01

    This paper addresses certain key feasibility issues facing the application of SiC-matrix microencapsulated fuels for light water reactor application. Issues addressed are the irradiation stability of the SiC-based nano-powder ceramic matrix under LWR-relevant irradiation conditions, the presence or extent of reaction of the SiC matrix with zirconium-based cladding, the stability of the inner and outer pyrolytic graphite layers of the TRISO coating system at this uncharacteristically low irradiation temperature, and the state of the particle–matrix interface following irradiation which could possibly affect thermal transport. In the process of determining these feasibility issues microstructural evolution and change in dimension and thermal conductivity was studied. As a general finding the SiC matrix was found to be quite stable with behavior similar to that of CVD SiC. In magnitude the irradiation-induced swelling of the matrix material was slightly higher and irradiation-degraded thermal conductivity was slightly lower as compared to CVD SiC. No significant reaction of this SiC-based nano-powder ceramic matrix material with Zircaloy was observed. Irradiation of the sample in the 320–360 °C range to a maximum dose of 7.7 × 10 25 n/m 2 (E > 0.1 MeV) did not have significant negative impact on the constituent layers of the TRISO coating system. At the highest dose studied, layer structure and interface integrity remained essentially unchanged with good apparent thermal transport through the microsphere to the surrounding matrix

  7. Adapting LWR to future needs: SECURE-P (PIUS)

    International Nuclear Information System (INIS)

    Hannerz, K.

    1984-01-01

    Advanced nuclear technology based on breeder reactors and fuel reprocessing may eventually be applied on a large scale, although the timing for this appears uncertain. However, in many parts of the world societal conditions and technological infrastructure mandate the use of a less complicated technology if the benefits of clean, safe nuclear power are to be available. Such a technology must be based on thermal reactors. Lack of fuel resources for their operation through most of the next century is unlikely to be a serious limitation. A natural contender would be the light water reactor, but today's designs lack many of the desired characteristics. However, introduction of certain new design features can eliminate the shortcomings and make the LWR the prime longterm candidate for a simple, technologically unsophisticated generation of nuclear power. Availability of such an option will also be a major asset for utilities in the large industrial countries before the advent of the era of advanced 'second generation' nuclear power. The costs of demonstrating the new design features are miniscule in relation to the benefits that should accrue. (author)

  8. Thermal conductivity of heterogeneous LWR MOX fuels

    Science.gov (United States)

    Staicu, D.; Barker, M.

    2013-11-01

    difference between these two bounds is 2% over the considered temperature range. The predictions obtained with the equations of Maxwell-Eucken (Eq. (3)) and Bergman (Eq. (4)) are equal and are in the interval between the series and parallel bounds. This result shows that the use of a sophisticated analytical or numerical model to predict the thermal conductivity is not justified [38]. The model of Maxwell-Eucken [31] was therefore chosen to predict the equivalent thermal conductivity of the heterogeneous MOX.The equivalent thermal conductivity of the stoichiometric heterogeneous MOX with an average PuO2 content of 7.2 wt.% (constituted by a stoichiometric UO2 matrix containing 15 vol.% of (U0.76Pu0.24)O1.975 agglomerates and 55 vol.% of a coating phase of (U0.94Pu0.06)O1.995) was calculated. The results show that the apparent thermal conductivity of the heterogeneous MOX, calculated using homogeneous MOX data (Eq. (5)) with O/M = 2.000, 1.995 and 1.975 (labeled Model 1 in Fig. 4) is not significantly different from the values measured by Duriez. The latter values are also very similar to the thermal conductivity of homogeneous MOX with O/M = 1.995. This simple model shows that the stoichiometry effect is sufficient to explain the lower thermal conductivity of LWR MOX fuel as compared to UO2. The advantage of this simple model is its consistency, as the calculations for the heterogeneous MOX are based on a unique formula for non-stoichiometric homogeneous (U,Pu)O2.In the second model, the effect of the plutonium is taken into account for the coating phase and for the Pu-rich agglomerates. The thermal conductivity is described by the correlations of Fink [16] for UO2.000, of Duriez et al. [2] for (U0.94Pu0.06)O1.995 (coating phase with low PuO2 content) and of Philipponneau [8] for (U0.76Pu0.24)O1.975 (Pu-rich agglomerates with high PuO2 content). The results (labeled Model 2 in Fig. 5) show that the calculated thermal conductivity of the heterogeneous 'stoichiometric' MOX is

  9. Environmental development plan. LWR commercial waste management

    International Nuclear Information System (INIS)

    1980-08-01

    This Environmental Development Plan (EDP) identifies the planning and managerial requirements and schedules needed to evaluate and assess the environmental, health and safety (EH and S) aspects of the Commercial Waste Management Program (CWM). Environment is defined in its broadest sense to include environmental, health (occupational and public), safety, socioeconomic, legal and institutional aspects. This plan addresses certain present and potential Federal responsibilities for the storage, treatment, transfer and disposal of radioactive waste materials produced by the nuclear power industry. The handling and disposal of LWR spent fuel and processed high-level waste (in the event reprocessing occurs) are included in this plan. Defense waste management activities, which are addressed in detail in a separate EDP, are considered only to the extent that such activities are common to the commercial waste management program. This EDP addresses three principal elements associated with the disposal of radioactive waste materials from the commercial nuclear power industry, namely Terminal Isolation Research and Development, Spent Fuel Storage and Waste Treatment Technology. The major specific concerns and requirements addressed are assurance that (1) radioactivity will be contained during waste transport, interim storage or while the waste is considered as retrievable from a repository facility, (2) the interim storage facilities will adequately isolate the radioactive material from the biosphere, (3) the terminal isolation facility will isolate the wastes from the biosphere over a time period allowing the radioactivity to decay to innocuous levels, (4) the terminal isolation mode for the waste will abbreviate the need for surveillance and institutional control by future generations, and (5) the public will accept the basic waste management strategy and geographical sites when needed

  10. Feasibility study on the development of advanced LWR fuel technology

    International Nuclear Information System (INIS)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others.

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO 2 pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO 2 -Gd 2 O 3 burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs

  11. Modelling of a LWR open fuel cycle using the message

    Energy Technology Data Exchange (ETDEWEB)

    Estanislau, Fidéllis B.G.L. e; Jonusan, Raoni A.S.; Costa, Antonella L.; Pereira, Claubia, E-mail: fidellis01@hotmail.com, E-mail: rjonusan@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    The main goal of the national energy planning is the development of a short and long-term strategies based on a holistic evaluation of all available energy sources guiding trends and delimiting expansion alternatives in the energetic sector. For a better understanding of the future possibilities, energy systems analyses are indispensable and support in the decision making related to the long term strategy and energy planning. Due to the projections for increased energy consumption according to the Energy Decennial Plan (year 2015) and the need to reduce greenhouse gas emissions presented by Brazil in the UNFCCC (United Nations Framework Convention on Climate Change), alternative energy sources such as solar, wind, nuclear and biomass sources have played an important role in the world energy matrix. In this way, since the nuclear energy is an option for the national energy mix, the present work aims to use the modelling tool MESSAGE (Model for Energy Supply System Alternatives and Their General Environmental Impact) to analyze and evaluate a nuclear power plant in an energy system. This tool is an optimization model for medium and long-term energy planning taking into account conversion and distribution technologies, energy policies and scenarios to satisfy a determined demand and systems constraints. In this work, a reproduction of results considering an LWR (Light Water Reactor) open-cycle are presented using a model in the MESSAGE code. (author)

  12. Validation of the LWR-EIR methods for the evaluation of compact beds

    International Nuclear Information System (INIS)

    Foskolos, K.; Grimm, P.; Maeder, C.; Paratte, J.M.

    1983-10-01

    The EIR code system for the calculation of light water reactors is presented and the methods used are briefly described. The application of the system on various types of critical experiments and benchmark problems proves its good precision, even for heterogeneous configurations with strong neutron absorbers like Boral. As the accuracy of the multiplication factor ksub(eff) is always better than 0.5% for normal LWR configurations, this code system is validated for the calculation of such configurations with a safety margin of 1.5% on ksub(eff). (Auth.)

  13. The need for LWR metrology standardization: the imec roughness protocol

    Science.gov (United States)

    Lorusso, Gian Francesco; Sutani, Takumichi; Rutigliani, Vito; van Roey, Frieda; Moussa, Alain; Charley, Anne-Laure; Mack, Chris; Naulleau, Patrick; Constantoudis, Vassilios; Ikota, Masami; Ishimoto, Toru; Koshihara, Shunsuke

    2018-03-01

    As semiconductor technology keeps moving forward, undeterred by the many challenges ahead, one specific deliverable is capturing the attention of many experts in the field: Line Width Roughness (LWR) specifications are expected to be less than 2nm in the near term, and to drop below 1nm in just a few years. This is a daunting challenge and engineers throughout the industry are trying to meet these targets using every means at their disposal. However, although current efforts are surely admirable, we believe they are not enough. The fact is that a specification has a meaning only if there is an agreed methodology to verify if the criterion is met or not. Such a standardization is critical in any field of science and technology and the question that we need to ask ourselves today is whether we have a standardized LWR metrology or not. In other words, if a single reference sample were provided, would everyone measuring it get reasonably comparable results? We came to realize that this is not the case and that the observed spread in the results throughout the industry is quite large. In our opinion, this makes the comparison of LWR data among institutions, or to a specification, very difficult. In this paper, we report the spread of measured LWR data across the semiconductor industry. We investigate the impact of image acquisition, measurement algorithm, and frequency analysis parameters on LWR metrology. We review critically some of the International Technology Roadmap for Semiconductors (ITRS) metrology guidelines (such as measurement box length larger than 2μm and the need to correct for SEM noise). We compare the SEM roughness results to AFM measurements. Finally, we propose a standardized LWR measurement protocol - the imec Roughness Protocol (iRP) - intended to ensure that every time LWR measurements are compared (from various sources or to specifications), the comparison is sensible and sound. We deeply believe that the industry is at a point where it is

  14. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  15. 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.

  16. Spent fuel data base: commercial light water reactors

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  17. Scaling analysis in bepu licensing of LWR

    Energy Technology Data Exchange (ETDEWEB)

    D' auria, Francesco; Lanfredini, Marco; Muellner, Nikolaus [University of Pisa, Pisa (Italy)

    2012-08-15

    'Scaling' plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a 'roadmap' to scaling. Key elements are the 'scaling-pyramid', related 'scaling bridges' and a logical path across scaling achievements (which constitute the 'scaling puzzle'). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the 'key-to-scaling', in the licensing process of a nuclear power plant. The proposed 'road map to scaling' aims at solving the 'scaling puzzle', by introducing a unified approach to the problem.

  18. Scaling analysis in bepu licensing of LWR

    International Nuclear Information System (INIS)

    D'auria, Francesco; Lanfredini, Marco; Muellner, Nikolaus

    2012-01-01

    'Scaling' plays an important role for safety analyses in the licensing of water cooled nuclear power reactors. Accident analyses, a sub set of safety analyses, is mostly based on nuclear reactor system thermal hydraulics, and therefore based on an adequate experimental data base, and in recent licensing applications, on best estimate computer code calculations. In the field of nuclear reactor technology, only a small set of the needed experiments can be executed at a nuclear power plant; the major part of experiments, either because of economics or because of safety concerns, has to be executed at reduced scale facilities. How to address the scaling issue has been the subject of numerous investigations in the past few decades (a lot of work has been performed in the 80thies and 90thies of the last century), and is still the focus of many scientific studies. The present paper proposes a 'roadmap' to scaling. Key elements are the 'scaling-pyramid', related 'scaling bridges' and a logical path across scaling achievements (which constitute the 'scaling puzzle'). The objective is addressing the scaling issue when demonstrating the applicability of the system codes, the 'key-to-scaling', in the licensing process of a nuclear power plant. The proposed 'road map to scaling' aims at solving the 'scaling puzzle', by introducing a unified approach to the problem.

  19. Location Based Services and Applications

    OpenAIRE

    Elenis Gorrita Michel; Rónier Sierra Dávila; Samuel Montejo Sánchez

    2012-01-01

    Location Based Services (LBS) continue to grow in popularity, effectiveness and reliability, to the extent that applications are designed and implemented taking into account the facilities of the user location information. In this work, some of the main applications are addressed, in order to make an assessment of the current importance of the LBS, as a branch of technology in full swing. In addition, the main techniques for location estimation are studied, essential information to the LBS. B...

  20. Preliminary concepts for detecting national diversion of LWR spent fuel

    International Nuclear Information System (INIS)

    Sonnier, C.S.; Cravens, M.N.

    1978-04-01

    Preliminary concepts for detecting national diversion of LWR spent fuel during storage, handling and transportation are presented. Principal emphasis is placed on means to achieve timely detection by an international authority. This work was sponsored by the Department of Energy/Office of Safeguards and Security (DOE/OSS) as part of the overall Sandia Fixed Facility Physical Protection Program

  1. Safety criteria related to microheterogeneities in LWR mixed oxide fuels

    International Nuclear Information System (INIS)

    Renard, A.; Mostin, N.

    1978-01-01

    The main safety aspets of PuO 2 microheterogeneities in the pellets of LWR mixed oxide fuels are reviewed. Points of interest are studied, especially the transient behaviour in accidental conditions and criteria are deduced for use in the specification and quality control of the fabricated product. (author)

  2. Nondestructive evaluation of LWR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructive testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined

  3. Contributions to LWR spent fuel storage and transport

    International Nuclear Information System (INIS)

    The papers included in this document describe the aspects of spent LWR fuel storage and transport-behaviour of spent fuel during storage; use of compact storage packs; safety of storage; design of storage facilities AR and AFR; description of transport casks and transport procedures

  4. BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) - Generation Methodology and Preliminary Testing of two ENEA-Bologna Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

    Science.gov (United States)

    Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela

    2016-02-01

    Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.

  5. A Stochastic LWR Model with Consideration of the Driver's Individual Property

    International Nuclear Information System (INIS)

    Tang Tieqiao; Wang Yunpeng; Yu Guizhen; Huang Haijun

    2012-01-01

    In this paper, we develop a stochastic LWR model based on the influences of the driver's individual property on his/her perceived density and speed deviation. The numerical results show that the driver's individual property has great effects on traffic flow only when the initial density is moderate, i.e., at this time, oscillating traffic flow will occur and the oscillating phenomena in the traffic system consisting of the conservative and aggressive drivers is more serious than that in the traffic system consisting of the conservative (aggressive) drivers.

  6. Review of tellurium release rates from LWR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Beahm, E.C.; Wichner, R.P.

    1983-01-01

    Although fission product tellurium presents a potentially significant radiohazard, its release and transport in source-term experiments is frequently overlooked because it does not possess a readily measurable, gamma emission; moreover, a recent study emphasized noble gas, iodine and cesium release from LWR fuel elements because of the large data base that exists for these materials. Some new tests show that in some cases tellurium may be held up in core material to a greater degree than previously assumed - an observation that prompts a careful reappraisal of the existing tellurium-release data and its chemical foundation

  7. Perspective on US NRC Policy Issues Concerning Use of Risk Insights for Non-LWR

    International Nuclear Information System (INIS)

    Ha, Jun Su; Kim, In Goo; Huh, Chang Wook; Kim, Kyun Tae

    2011-01-01

    Since the PRA Implementation plan of US NRC (1994), PRA has been applied to all NPPs in USA and risk insights have been used for the regulation as a complement of the deterministic approaches. RIRIP (Risk-Informed Regulation Implementation Plan, 2000) and RPP (Risk-Informed and Performance-Based Plan, 2007) were announced by US NRC thereafter, which recommended enhanced use of risk insights. In the meantime, there have been lots of policy issues concerning use of risk insights for licensing Non-LWR designs, which will be discussed in this paper to understand the stream of perspectives on US NRC's approach

  8. Risk management: integration of social and technical risk variables into safety assessments of LWR'S

    International Nuclear Information System (INIS)

    Turnage, J.J.; Husseiny, A.A.

    1980-01-01

    A risk management methodology is developed here to formalize the acceptability levels of commercial LWR power plants via the estimation of risk levels acceptable to the public and the integration of such estimates into risk-benefit analysis. Utility theory is used for developing preference models based on value trade-offs among multiple objectives and uncertainties about the impact of alternatives. The method involves reducing the various variables affecting safety acceptability decisions to a single function that provides a metric for acceptability levels. The function accomondates for technical criteria related to design and licensing decisions, as well as public reactions to certain choices

  9. The necessity of improvement for the current LWR fuel assembly homogenization method

    International Nuclear Information System (INIS)

    Tang Chuntao; Huang Hao; Zhang Shaohong

    2007-01-01

    When the modern LWR core analysis method is used to do core nuclear design and in-core fuel management calculation, how to accurately obtain the fuel assembly homogenized parameters is a crucial issue. In this paper, taking the NEA C5G7-MOX benchmark problem as a severe test problem, which involves low-enriched uranium assemblies interspersed with MOX assemblies, we have re-examined the applicability of the two major assumptions of the modern equivalence theory for fuel assembly homoge- nization, i.e. the isolated assembly spatial spectrum assumption and the condensed two- group representation assumption. Numerical results have demonstrated that for LWR cores with strong spectrum interaction, both of these two assumptions are no longer applicable and the improvement for the homogenization method is necessary, the current two-group representation should be improved by the multigroup representation and the current reflective assembly boundary condition should be improved by the 'real' assembly boundary condition. This is a research project supported by National Natural Science Foundation of China (10605016). (authors)

  10. Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

    International Nuclear Information System (INIS)

    Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

    2001-01-01

    The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

  11. Survey of LWR environmental control technology performance and cost

    International Nuclear Information System (INIS)

    Heeb, C.M.; Aaberg, R.L.; Cole, B.M.; Engel, R.L.; Kennedy, W.E. Jr.; Lewallen, M.A.

    1980-03-01

    This study attempts to establish a ranking for species that are routinely released to the environment for a projected nuclear power growth scenario. Unlike comparisons made to existing standards, which are subject to frequent revision, the ranking of releases can be used to form a more logical basis for identifying the areas where further development of control technology could be required. This report describes projections of releases for several fuel cycle scenarios, identifies areas where alternative control technologies may be implemented, and discusses the available alternative control technologies. The release factors were used in a computer code system called ENFORM, which calculates the annual release of any species from any part of the LWR nuclear fuel cycle given a projection of installed nuclear generation capacity. This survey of fuel cycle releases was performed for three reprocessing scenarios (stowaway, reprocessing without recycle of Pu and reprocessing with full recycle of U and Pu) for a 100-year period beginning in 1977. The radioactivity releases were ranked on the basis of a relative ranking factor. The relative ranking factor is based on the 100-year summation of the 50-year population dose commitment from an annual release of radioactive effluents. The nonradioactive releases were ranked on the basis of dilution factor. The twenty highest ranking radioactive releases were identified and each of these was analyzed in terms of the basis for calculating the release and a description of the currently employed control method. Alternative control technology is then discussed, along with the available capital and operating cost figures for alternative control methods

  12. Feasibility study on the development of advanced LWR fuel technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Youn Ho; Sohn, D. S.; Jeong, Y. H.; Song, K. W.; Song, K. N.; Chun, T. H.; Bang, J. G.; Bae, K. K.; Kim, D. H. and others

    1997-07-01

    Worldwide R and D trends related to core technology of LWR fuels and status of patents have been surveyed for the feasibility study. In addition, various fuel cycle schemes have been studied to establish the target performance parameters. For the development of cladding material, establishment of long-term research plan for alloy development and optimization of melting process and manufacturing technology were conducted. A work which could characterize the effect of sintering additives on the microstructure of UO{sub 2} pellet has been experimentally undertaken, and major sintering variables and their ranges have been found in the sintering process of UO{sub 2}-Gd{sub 2}O{sub 3} burnable absorber pellet. The analysis of state of the art technology related to flow mixing device for spacer grid and debris filtering device for bottom nozzle and the investigation of the physical phenomena related to CHF enhancement and the establishment of the data base for thermal-hydraulic performance tests has been done in this study. In addition, survey on the documents of the up-to-date PWR fuel assemblies developed by foreign vendors have been carried out to understand their R and D trends and establish the direction of R and D for these structural components. And, to set the performance target of the new fuel, to be developed, fuel burnup and economy under the extended fuel cycle length scheme were estimated. A preliminary study on the failure mechanism of CANDU fuel, key technology and advanced coating has been performed. (author). 190 refs., 31 tabs., 129 figs.

  13. Safety aspects and operating experience of LWR plants in Japan

    International Nuclear Information System (INIS)

    Aoki, S.; Hinoki, M.

    1977-01-01

    From the outset of nuclear power development in Japan, major emphasis has been placed on the safety of the nuclear power plants. There are now twelve nuclear power plants in operation with a total output of 6600 MWe. Their operating records were generally satisfactory, but in the 1974 to 1975 period, they experienced somewhat declined availability due to the repair work under the specific circumstances. After investigation of causes of troubles and the countermeasures thereof were made to ensure safety, they are now keeping good performance. In Japan, nuclear power plants are strictly subject to sufficient and careful inspection in compliance with the safety regulation, and are placed under stringent radiation control of employees. Under the various circumstances, however, the period of annual inspection tends to be prolonged more than originally planned, and this consequently is considered to be one of the causes of reduced availability. In order to develop nuclear power generation for the future, it is necessary to put further emphasis on the assurance of safety and to endeavor to devise measures to improve availability of the plants, based on the careful analysis of causes which reduce plant availability. This paper discusses the results of studies made for the following items from such viewpoints: (1) Safety and Operating Experience of LWR Nuclear Power Plants in Japan; a) Operating experience with light water reactors b) Improvements in design of light water reactors during the past ten years c) Analysis of the factors which affect plant availability; 2) Assurance of Safety and Measures to Increase Availability a) Measures for safety and environmental protection b) Measures to reduce radiation exposure of employees c) Appropriateness of maintenance and inspection work d) Measures to increase plant availability e) Measures to improve reliability of equipments and components; and 3) Future Technical Problems

  14. A review on future trends of LWR fuel cycle costs

    International Nuclear Information System (INIS)

    Tamiya, S.; Otomo, T.; Meguro, T.

    1977-01-01

    In the cost estimations in the past, the main components of fuel cycle were mining and milling, uranium enrichment and fuel fabrication, and reprocessing charge deemed to be recovered by plutonium credit. Since the oil crisis, every component of fuel cycle cost has gone up in recent years as well as the construction cost of a power station. Recent analysis shows that the costs in the back end of fuel cycle are much higher than those anticipated several years ago, although their contribution to the electricity generating cost by nuclear would be small. The situation of the back end of the fuel cycle has been quite changed in recent years, and there are still many uncertainties in this field, that is, regulatory requirements for reprocessing plant such as safety, safeguards, environmental protection and high level waste management. So, it makes it more difficult to estimate the investment in this sector of fuel cycle, therefore, to estimate the cost of this sector. The institutional problems must be cleared in relation to the ultimate disposal of high level waste, too. Co-location of some parts of fuel cycle facilities may also affect on the fuel cycle costs. In this paper a review is made of the future trend of nuclear fuel cycle cost of LWR based on the recent analysis. Those factors which affect the fuel cycle costs are also discussed. In order to reduce the uncertainties of the cost estimations as soon as possible, the necessity is emphasized to discuss internationally such items as the treatment and disposal of high level radioactive wastes, siting issues of a reprocessing plant, physical protection of plutonium and the effects of plutonium on the environment

  15. EPICS Application Based on RTEMS

    International Nuclear Information System (INIS)

    Shi Haoli; Wang Chunhong; Tang Jingyu

    2009-01-01

    At present, in the accelerator field all over the world, the control system is established with EPICS, which is a configuration software package based on Ethernet and distributed control system. Early version of EPICS was developed based on VxWorks. Now, EPICS international collaboration group is dedicating to supporting open source, free RTEMS, and extending RTEMS into the newer version of EPICS3.14. The paper illustrates the development, characteristic of RTEMS, compares the performances of several ordinary RTOS, builds RTEMS operating system based on MVME5500, and finally presents an EPICS application example on RTEMS. (authors)

  16. Development of advanced LWR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H. [and others

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out.

  17. Development of advanced LWR fuel cladding

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan; Park, S. Y.; Lee, M. H.

    2000-04-01

    This report describes the results from evaluating the preliminary Zr-based alloys to develop the advanced Zr-based alloys for the nuclear fuel claddings, which should have good corrosion resistance and mechanical properties at high burn-up over 70,000MWD/MTU. It also includes the results from the basic studies for optimizing the processes which are involved in the development of the advanced Zr-based alloys. Ten(10) kinds of candidates for the alloys of which performance is over that of the existing Zircaloy-4 or ZIRLO alloy were selected out of the preliminary alloys of 150 kinds which were newly designed and repeatedly manufactured and evaluated to find out the promising alloys. First of all, the corrosion tests on the preliminary alloys were carried out to evaluate their performance in both pure water and LiOH solution at 360 deg C and in steam at 400 deg C. The tensile tests were performed on the alloys which proved to be good in the corrosion resistance. The creep behaviors were tested at 400 deg C for 10 days with the application of constant load on the samples which showed good performance in the corrosion resistance and tensile properties. The effect of the final heat treatment and A-parameters as well as Sn or Nb on the corrosion resistance, tensile properties, hardness, microstructures of the alloys was evaluated for some alloys interested. The other basic researches on the oxides, electrochemical properties, corrosion mechanism, and the establishment of the phase diagrams of some alloys were also carried out

  18. Estimation of the core-wide fuel rod damage during a LWR LOCA

    International Nuclear Information System (INIS)

    Mattila, L.; Sairanen, R.; Stengaard, J.-O.

    1975-01-01

    The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

  19. Assesment On The Possibility To Modify Fabrication Equipment For Fabrication Of HWR And LWR Fuel Elements

    International Nuclear Information System (INIS)

    Tri-Yulianto

    1996-01-01

    Based on TOR BATAN for PELITA VI. On of BATAN program in the fuel element production technology section is the acquisition of the fuel element fabrication technology for research reactor as well as power reactor. The acquisition can be achieved using different strategies, e.g. by utilizing the facility owned for research and development of the technology desired or by transferring the technology directly from the source. With regards to the above, PEBN through its facility in BEBE has started the acquisition of the fuel element fabrication technology for power reactor by developing the existing equipment initially designed to fabricate HWR Cinere fuel element. The development, by way of modifying the equipment, is intended for the production of HWR (Candu) and LWR (PWR and BWR) fuel elements. To achieve above objective, at the early stage of activity, an assesment on the fabrication equipment for pelletizing, component production and assembly. The assesment was made by comparing the shape and the size of the existing fuel element with those used in the operating reactors such as Candu reactors, PWR and BWR. Equipment having the potential to be modified for the production of HWR fuel elements are as followed: For the pelletizing equipment, the punch and dies can be used of the pressing machine for making green pellet can be modified so that different sizes of punch and dies can be used, depending upon the size of the HWR and LWR pellets. The equipment for component production has good potential for modification to produce the HWR Candu fuel element, which has similar shape and size with those of the existing fuel element, while the possibility of producing the LWR fuel element component is small because only a limited number of the required component can be made with the existing equipment. The assembly equipment has similar situation whit that of the component production, that is, to assemble the HWR fuel element modification of few assembly units very probable

  20. PLUTON: A Three-Group Model for the Radial Distribution of Plutonium, Burnup, and Power Profiles in Highly Irradiated LWR Fuel Rods

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe

    2001-01-01

    A three-group model (PLUTON) is described, which predicts the power density distribution, plutonium buildup, and burnup profiles across the fuel pellet radius as a function of in-pile time and parameters characterizing the type of reactor system with respect to fuel temperature and changes of density during the irradiation period. The PLUTON model is a part of two fuel performance codes (ASFAD and FEMAXI-V), which provide all necessary input for this model, mainly local temperatures and fuel matrix density across the radius. Comparisons between measurements and predictions of the PLUTON model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnup between 21 000 and 64 000 MWd/t. It is shown that the PLUTON predictions are in good agreement with measurements as well as with predictions of the well-known TUBRNP model. The proposed model is flexibly applicable for all types of light water reactor (LWR) fuels, including mixed oxide, and for fuel tested in the Organization for Economic Corporation and Development's Halden heavy water reactor. The PLUTON three-group model is based on analytical (theoretical) consideration of neutron absorption in a resonant region of the fuel in its apparent form. It makes the model more flexible in comparison with the semi-empirical TUBRNP one-group model and allows the physically based model analysis of commercial LWR-type fuels at high burnup as well as analysis of experimental fuel rods tested in the Halden heavy water reactor, which is one of the main test reactors in the world. The differences in fuel behavior in the Halden reactor in terms of burnup distribution and plutonium buildup can be more clearly understood with the PLUTON model

  1. Engineered zircaloy cladding modifications for improved accident tolerance of LWR fuel: US DOE NEUP Integrated Research Project

    International Nuclear Information System (INIS)

    Heuser, Brent

    2013-01-01

    An integrated research project (IRP) to fabricate and evaluate modified zircaloy LWR cladding under normal BWR/PWR operation and off-normal events has been funded by the US DOE. The IRP involves three US academic institutions, a US national laboratory, an intermediate stock industrial cladding supplier, and an international academic institution. A combination of computational and experimental protocols will be employed to design and test modified zircaloy cladding with respect to corrosion and accelerated oxide growth, the former associated with normal operation, the latter associated with steam exposure during loss of coolant accidents (LOCAs) and low-pressure core re-floods. Efforts will be made to go beyond design-base accident (DBA) scenarios (cladding temperature equal to or less than 1204 deg. C) during the experimental phase of modified zircaloy performance characterisation. The project anticipates the use of the facilities at ORNL to achieve steam exposure beyond DBA scenarios. In addition, irradiation of down-selected modified cladding candidates in the ATR may be performed. Cladding performance evaluation will be incorporated into a reactor system modelling effort of fuel performance, neutronics, and thermal hydraulics, thereby providing a holistic approach to accident-tolerant nuclear fuel. The proposed IRP brings together personnel, facilities, and capabilities across a wide range of technical areas relevant to the study of modified nuclear fuel and LWR performance during normal operation and off-normal scenarios. Two pathways towards accident-tolerant LWR fuel are envisioned, both based on the modification of existing zircaloy cladding. The first is the modification of the cladding surface by the application of a coating layer designed to shift the M + O→MO reaction away from oxide growth during steam exposure at elevated temperatures. This pathway is referred to as the 'surface coating' solution. The second is the modification of the bulk

  2. Summary of aerosol code-comparison results for LWR aerosol containment tests LA1, LA2, and LA3

    International Nuclear Information System (INIS)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1987-01-01

    The light-water reactor (LWR) aerosol containment experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities for the LACE tests are being coordinated at the Oak Ridge National Laboratory. For each of the six experiments, pretest calculations (for code-to-code comparisons) and blind post-test calculations (for code-to-test data comparisons) are being performed. This paper presents a summary of the pretest aerosol-code results for tests LA1, LA2, and LA3

  3. Improving the safety of LWR power plants. Final report

    International Nuclear Information System (INIS)

    1980-04-01

    This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs

  4. Development of top nozzle for Korean standard LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S. K.; Kim, I. K.; Choi, K. S.; Kim, Y. H.; Lee, J. N.; Kim, H. K. [KNFC, Taejon (Korea, Republic of)

    2001-10-01

    Performance evaluation was executed for each component and its assembly for the deduced Top Nozzles to develop the new Top Nozzle for LWR. This new Top Nozzle is composed of the optimum components among the derived Top Nozzles that have been evaluated in the viewpoint of structural integrity, simpleness of dismantle and assembly, manufacturability etc. In this study, the developed Top Nozzle satisfied all the related design criteria. In special, it makes fuel repair time reduced by assembling and disassembling itself as one body, and improves Fuel Assembly holddown ability by revising the design parameters of its spring and the structural integrity through the betterment of its geometrical shpae of Flange and Holddown Plate as compared with the existing LWR Top Nozzles.

  5. Development of LWR fuel performance code FEMAXI-6

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2006-01-01

    LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

  6. Review and comparison of WWER and LWR Codes and Standards

    International Nuclear Information System (INIS)

    Buckthorpe, D.; Tashkinov, A.; Brynda, J.; Davies, L.M.; Cueto-Felgeueroso, C.; Detroux, P.; Bieniussa, K.; Guinovart, J.

    2003-01-01

    The results of work on a collaborative project on comparison of Codes and Standards used for safety related components of the WWER and LWR type reactors is presented. This work was performed on behalf of the European Commission, Working Group Codes and Standards and considers areas such as rules, criteria and provisions, failure mechanisms , derivation and understanding behind the fatigue curves, piping, materials and aging, manufacturing and ISI. WWERs are essentially designed and constructed using the Russian PNAE Code together with special provisions in a few countries (e.g. Czech Republic) from national standards. The LWR Codes have a strong dependence on the ASME Code. Also within Western Europe other codes are used including RCC-M, KTA and British Standards. A comparison of procedures used in all these codes and standards have been made to investigate the potential for equivalencies between the codes and any grounds for future cooperation between eastern and western experts in this field. (author)

  7. Processing of the GALILEO fuel rod code model uncertainties within the AREVA LWR realistic thermal-mechanical analysis methodology

    International Nuclear Information System (INIS)

    Mailhe, P.; Barbier, B.; Garnier, C.; Landskron, H.; Sedlacek, R.; Arimescu, I.; Smith, M.; Bellanger, P.

    2013-01-01

    The availability of reliable tools and associated methodology able to accurately predict the LWR fuel behavior in all conditions is of great importance for safe and economic fuel usage. For that purpose, AREVA has developed its new global fuel rod performance code GALILEO along with its associated realistic thermal-mechanical analysis methodology. This realistic methodology is based on a Monte Carlo type random sampling of all relevant input variables. After having outlined the AREVA realistic methodology, this paper will be focused on the GALILEO code benchmarking process, on its extended experimental database and on the GALILEO model uncertainties assessment. The propagation of these model uncertainties through the AREVA realistic methodology is also presented. This GALILEO model uncertainties processing is of the utmost importance for accurate fuel design margin evaluation as illustrated on some application examples. With the submittal of Topical Report GALILEO to the U.S. NRC in 2013, GALILEO and its methodology are on the way to be industrially used in a wide range of irradiation conditions. (authors)

  8. Equipment designs for the spent LWR fuel dry storage demonstration

    International Nuclear Information System (INIS)

    Steffen, R.J.; Kurasch, D.H.; Hardin, R.T.; Schmitten, P.F.

    1980-01-01

    In conjunction with the Spent Fuel Handling and Packaging Program (SFHPP) equipment has been designed, fabricated and successfully utilized to demonstrate the packaging and interim dry storage of spent LWR fuel. Surface and near surface storage configurations containing PWR fuel assemblies are currently on test and generating baseline data. Specific areas of hardware design focused upon include storage cell components and the support related equipment associated with encapsulation, leak testing, lag storage, and emplacement operations

  9. Safety-related LWR research. Annual report 1993

    International Nuclear Information System (INIS)

    Hueper, R.

    1994-06-01

    The reactor safety R and D work of the Karlsruhe Nuclear Research Centre (KfK) has been part of the Nuclear Safety Research Project (PSF) since 1990. The present annual report 1993 summarizes the results on LWR safety. The research tasks are coordinated in agreement with internal and external working groups. The contributions to this report correspond to the status at the end of 1993. (orig./HP) [de

  10. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    International Nuclear Information System (INIS)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables

  11. Baseline descriptions for LWR spent fuel storage, handling, and transportation

    Energy Technology Data Exchange (ETDEWEB)

    Moyer, J.W.; Sonnier, C.S.

    1978-04-01

    Baseline descriptions for the storage, handling, and transportation of reactor spent fuel are provided. The storage modes described include light water reactor (LWR) pools, away-from-reactor basins, dry surface storage, reprocessing-facility interim storage pools, and deep geologic storage. Land and water transportation are also discussed. This work was sponsored by the Department of Energy/Office of Safeguards and Security as part of the Sandia Laboratories Fixed Facility Physical Protection Program. 45 figs, 4 tables.

  12. Standard casks for the transport of LWR spent fuel

    International Nuclear Information System (INIS)

    Blum, P.

    1986-01-01

    During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufactured in different countries and are presently used for european and intercontinental transports. The main advantages of these casks are: large payload, moderate cost, reliability, standardisation facilitating fabrication, operation and spare part supply [fr

  13. Investigation of valve failure problems in LWR power plants

    International Nuclear Information System (INIS)

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

  14. Modular approach to LWR in-core fuel management

    International Nuclear Information System (INIS)

    Urli, N.; Pevec, D.; Coffou, E.; Petrovic, B.

    1980-01-01

    The most important methods in the LWR in-core fuel management are reviewed. A modular approach and optimization by use of infinite multiplication factor and power form-factor are favoured. A computer program for rotation of fuel assemblies at reloads has been developed which improves further fuel economy and reliability of nuclear power plants. The program has been tested on the PWR core and showed to decrease the power form-factors and flatten the radial power distribution. (author)

  15. Development of information management system on LWR spent fuel

    International Nuclear Information System (INIS)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S.

    2002-01-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility

  16. Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Geun Hyeong; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    LWR uses fuel as {sup 235}U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile {sup 233}U when {sup 232}Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster.

  17. Preliminary Study for Radioactivity Evaluation of MSR compared with LWR

    International Nuclear Information System (INIS)

    Lee, Geun Hyeong; Kim, Hee Reyoung

    2014-01-01

    LWR uses fuel as 235 U and fissile material as solid (enriched uranium). Those cannot control its component artificially and hard to change fuel frequently. Therefore this fuel remains as much as possible. That makes risk of high radiation leakage because of long neutron irradiation time. On the other hand, MSR (Molten Salt Reactor) uses fuel as thorium-uranium; fissile 233 U when 232 Th absorbs one neutron, and fissile material as liquid (molten salt). It has plenty of benefits respect to radioactive safety. It leads nuclear fuel dump when accident happens, diminishes basic fission substances' radiation and even the cost (Th exist 3∼4 times more on the earth compared with natural uranium). Source term is much lower than conventional LWR in order to processing time. Radiation exposure from volatile fission products in severe accidents is thought to be negligible due to the continuous removal mechanism. The generation of high level radioactive wastes from MSR is estimated to be much smaller than that of conventional LWR because of its less converting probability of thorium to minor actinides. It was thought the fundamental approach to MSR would make it possible to realize the safety of reactor when considering the severe accidents affecting on nuclear power plants due to natural disaster

  18. Energy profit ratio on LWR by uranium recycles

    International Nuclear Information System (INIS)

    Amano, Osamu; Uno, Takeki; Matsushima, Jun

    2009-01-01

    Energy profit ratio is defined as the ratio of output energy/input system total energy. In case of electric power generation, input energy is a total for fuel such as uranium mining and enrichment, fuel transportation, build nuclear power plant, M and O and for disposal waste and decommission of reactor vessel. Output energy is the total electricity on LWR during the plant life. EPR on both PWR and BWR is high value using gas centrifuge enrichment compared other type of electric power generation such as a thermal power, a hydraulic power, a wind power and a photovoltaic power. How is the EPR on LWR by MOX? We need understanding the energy of reprocessing spent fuel, MOX fuel fabrication, low level waste disposal and high level radioactive glass disposal. As we show the material balance for two cases, the first is the case of long term storage and reprocessing before FBR, the second is the MOX fuel cycle on LWR plant. The MOX fuel recycle is better EPR value rather than the case of long term storage and reprocessing before FBR (LTSRBF). At the gaseous diffusion enrichment case, MOX fuel recycle has 15 to 18% higher EPR value than LTSRBF. At the gas centrifuge enrichment case the MOX fuel recycle has 17 to 18 higher EPR value than LTSRBF. MOX fuel recycle decreases the uranium mining and refine mass, enrichment separative work and the spent fuel interim storage. It tells us the MOX fuel recycle is good way from view of EPR. (author)

  19. Development of information management system on LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. D.; Lee, S. H.; Song, D. Y.; Jeon, I.; Park, S. J.; Seo, D. S. [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    LWRs in Korea should manage all the information of spent fuel to implement the obligations under Korea-IAEA safeguards agreement and to perform the nuclear material accountancy work at the facility level. The information management system on LWR spent fuel was developed to manage all movement records from receipt to shipment of LWR fuels, and to get the necessary information such as nuclear fuel inventory lists and status, maps of fresh fuel storage, reactor and spent fuel pool, receipt and shipment records and so on. This information management system has a function to setup the system environments to cover the various kinds of storage types for all LWRs ; reactor, spent fuel pool and fresh fuel storage. The movements of nuclear fuel between the storages can be easily done by double click of the mouse to the destination. It also has a several error checking routines for maintaining the correct accounting data. Using this information management system of LWR spent fuel, facility operators can perform efficiently and effectively the safeguards related works including nuclear material accountancy at each facility.

  20. Simulated Fission Gas Behavior in Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, Kun [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States); Harp, Jason [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-15

    As a promising candidate for the accident tolerant fuel (ATF) used in light water reactors (LWRs), the fuel performance of uranium silicide (U3Si2) at LWR conditions needs to be well-understood. However, existing experimental post-irradiation examination (PIE) data are limited to the research reactor conditions, which involve lower fuel temperature compared to LWR conditions. This lack of appropriate experimental data significantly affects the development of fuel performance codes that can precisely predict the microstructure evolution and property degradation at LWR conditions, and therefore evaluate the qualification of U3Si2 as an AFT for LWRs. Considering the high cost, long timescale, and restrictive access of the in-pile irradiation experiments, this study aims to utilize ion irradiation to simulate the inpile behavior of the U3Si2 fuel. Both in situ TEM ion irradiation and ex situ high-energy ATLAS ion irradiation experiments were employed to simulate different types of microstructure modifications in U3Si2. Multiple PIE techniques were used or will be used to quantitatively analyze the microstructure evolution induced by ion irradiation so as to provide valuable reference for the development of fuel performance code prior to the availability of the in-pile irradiation data.

  1. Study on reprocessing plant during transition period from LWR to FBR

    International Nuclear Information System (INIS)

    Shimada, Takashi; Matsui, Minefumi; Nishimura, Masashi; Ishida, Yasuhiro; Mori, Yukihide; Kuroda, Kazuhiko

    2011-01-01

    We have proposed a concept of a reprocessing plant suitable for the transition period from the light water reactors (LWRs) to the fast breeder reactors (FBRs) by making comparison of two plant concepts: (1) Independent Plant which processes LWR fuel and FBR fuel in separately constructed lines and (2) Modularized Plant which processes LWR fuel and FBR fuel in a same line. We made construction plans based on the reference power generation plan, and evaluated the Pu supply capability using the power generation plan as an indicator of plant operation flexibility. In general, a margin of processing capacity increases the Pu supply capability. The margin of the Modularized Plant necessary to obtain equivalent Pu supply capability is smaller than that of the Independent Plant. Also the margin of the Independent Plant results in decrease in the plant utilization factor. But the margin of the Modularized Plant results in little decrease in the plant utilization factor, because the Modularized Plant can address the types of reprocessing fuel to adjust to Pu demand and processing capacity. Therefore, the Modularized Plant has a greater potential for the reprocessing plants during transition period. (author)

  2. Phoenix type concepts for transmutation of LWR waste minor actinides

    International Nuclear Information System (INIS)

    Segev, M.

    1994-01-01

    A number of variations on the original Phoenix theme were studied. The basic rationale of the Phoenix incinerator is making oxide fuel of the LWR waste minor actinides, loading it in an FFTF-like subcritical core, then bombarding the core with the high current beam accelerated protons to generate considerable energy through spallation and fission reactions. As originally assessed, if the machine is fed with 1600 MeV protons in a 102 mA current, then 8 core modules are driven to transmute the yearly minor actinides waste of 75 1000 MW LWRs into Pu 238 and fission products; in a 2 years cycle the energy extracted is 100000 MW d/T. This performance cannot be substantiated in a rigorous analysis. A calculational consistent methodology, based on a combined execution of the Hermes, NCNP, and Korigen codes, shows, nonetheless that changes in the original Phoenix parameters can upgrade its performance.The original Phoenix contains 26 tons minor actinides in 8 core modules; 1.15 m 3 module is shaped for 40% neutron leakage; with a beam of 102 mA the 8 modules are driven to 100000 MW/T in 10.5 years, burning out the yearly minor actinide waste of 15 LWRs; the operation must be assisted by grid electricity. If the 1.15 m 3 module is shaped to allow only 28% leakage, then a beam of 102 mA will drive the 8 modules to 100000 MW/T in 3.5 years, burning out the yearly minor actinides waste of 45 LWRs. Some net grid electricity will be generated. If 25 tons minor actinides are loaded into 5 modules, each 1.72 m 3 in volume and of 24% leakage, then a 97 mA beam will drive the module to 100000 MW/T in 2.5 years, burning out the yearly minor actinides waste of 70 LWRs. A considerable amount of net grid electricity will be generated. If the lattice is made of metal fuel, and 26 tons minor actinides are loaded into 32 small modules, 0.17 m 3 each, then a 102 mA beam will drive the modules to 100000 MW/T in 2 years, burning out the yearly minor actinides waste of 72 LWRs. A considerable

  3. Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

    Science.gov (United States)

    Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

    2018-02-01

    SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

  4. Extending dry storage of spent LWR fuel for up to 100 years

    International Nuclear Information System (INIS)

    Einziger, R.E.; McKinnon, M.A.; Machiels, A.J.

    1999-01-01

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and excessive

  5. Extending dry storage of spent LWR fuel for up to 100 years

    International Nuclear Information System (INIS)

    Einziger, R. E.

    1998-01-01

    Because of delays in closing the back end of the fuel cycle in the U.S., there is a need to extend dry inert storage of spent fuel beyond its originally anticipated 20-year duration. Many of the methodologies developed to support initial licensing for 20-year storage should be able to support the longer storage periods envisioned. This paper evaluates the applicability of existing information and methodologies to support dry storage up to 100 years. The thrust of the analysis is the potential behavior of the spent fuel. In the USA, the criteria for dry storage of LWR spent fuel are delineated in 10 CFR 72 [1]. The criteria fall into four general categories: maintain subcriticality, prevent the release of radioactive material above acceptable limits, ensure that radiation rates and doses do not exceed acceptable levels, and maintain retrievability of the stored radioactive material. These criteria need to be considered for normal, off-normal, and postulated accident conditions. The initial safety analysis report submitted for licensing evaluated the fuel's ability to meet the requirements for 20 years. It is not the intent to repeat these calculations, but to look at expected behavior over the additional 80 years, during which the temperatures and radiation fields are lower. During the first 20 years, the properties of the components may change because of elevated temperatures, presence of moisture, effects of radiation, etc. During normal storage in an inert atmosphere, there is potential for the cladding mechanical properties to change due to annealing or interaction with cask materials. The emissivity of the cladding could also change due to storage conditions. If there is air leakage into the cask, additional degradation could occur through oxidation in breached rods, which could lead to additional fission gas release and enlargement of cladding breaches. Air in-leakage could also affect cover gas conductivity, cladding oxidation, emissivity changes, and

  6. Least squares methodology applied to LWR-PV damage dosimetry, experience and expectations

    International Nuclear Information System (INIS)

    Wagschal, J.J.; Broadhead, B.L.; Maerker, R.E.

    1979-01-01

    The development of an advanced methodology for Light Water Reactors (LWR) Pressure Vessel (PV) damage dosimetry applications is the subject of an ongoing EPRI-sponsored research project at ORNL. This methodology includes a generalized least squares approach to a combination of data. The data include measured foil activations, evaluated cross sections and calculated fluxes. The uncertainties associated with the data as well as with the calculational methods are an essential component of this methodology. Activation measurements in two NBS benchmark neutron fields ( 252 Cf ISNF) and in a prototypic reactor field (Oak Ridge Pool Critical Assembly - PCA) are being analyzed using a generalized least squares method. The sensitivity of the results to the representation of the uncertainties (covariances) was carefully checked. Cross element covariances were found to be of utmost importance

  7. Measurement and characterization of fission products released from LWR fuel

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. This project was sponsored by the USNRC under a broad program of reactor safety studies. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from approx. 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag

  8. Conceptual design of a spent LWR fuel recycle complex

    International Nuclear Information System (INIS)

    Kirk, B.H.

    1980-01-01

    Purpose was to design a licensable facility, to make cost-benefit analyses of alternatives, and to aid in developing licensing criteria. The Savannah River Plant was taken to be the site for the recycle complex. The spent LWR fuel will be processed through the plant at the rate of 3000 metric tons of heavy metal per year. The following aspects of the complex are discussed: operation, maintenance, co-conversion (Coprecal), waste disposal, off-gas treatment, ventilation, safeguards, accounting, equipment and fuel fabrication. Differences between the co-processing case and the separated streams case are discussed. 44 figures

  9. Hamor-2: a computer code for LWR inventory calculation

    International Nuclear Information System (INIS)

    Guimaraes, L.N.F.; Marzo, M.A.S.

    1985-01-01

    A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

  10. Measurement and characterization of fission products released from LWR fuel

    International Nuclear Information System (INIS)

    Osborne, M.F.; Collins, J.L.; Lorenz, R.A.; Norwood, K.S.; Strain, R.V.

    1984-01-01

    Samples of commercial LWR fuel have been heated under simulated accident conditions to determine the extent and the chemical forms of fission product release. Of the five tests discussed, the fractional releases of Kr, I, and Cs varied from proportional 2% at 1400 0 C to >50% at 2000 0 C; much smaller fractions of Ru, Ag, Sb, and Te were measured in some tests. The major chemical forms in the effluent appeared to include CsI, CsOH, Sb, Te, and Ag. (orig./HP)

  11. Nondestructive evaluation of LWR spent fuel shipping casks

    International Nuclear Information System (INIS)

    Ballard, D.W.

    1978-02-01

    An analysis of nondestructve testing (NDT) methods currently being used to evaluate the integrity of Light Water Reactor (LWR) spent fuel shipping casks is presented. An assessment of anticipated NDT needs related to breeder reactor cask requirements is included. Specific R and D approaches to probable NDT problem areas such as the evaluation of austenitic stainless steel weldments are outlined. A comprehensive bibliography of current NDT methods for cask evaluation in the USA, Great Britain, Japan and West Germany was compiled for this study

  12. Standard casks for the transport of LWR spent fuel

    International Nuclear Information System (INIS)

    Blum, P.

    1985-01-01

    During the past decade, TRANSNUCLEAIRE has developed, licensed and marketed a family of standard casks for the transport of spent fuel from LWR reactors to reprocessing plants and the ancillary equipments necessary for their operation and transport. A large number of these casks have been manufacturer under TRANSNUCLEAIRE supervision in different countries and are presently used for European and intercontinental transports. The main advantages of these casks are: - large payload for considered modes of transport, - moderate cost, - reliability due to the large experience gained by TRANSNUCLEAIRE as concerns fabrication and operation problems, - standardisation facilitating fabrication, operation and spare part supply [fr

  13. A study for small-medium LWR development of JAPC

    International Nuclear Information System (INIS)

    Okazaki, Toshihiko; Hida, Takahiko; Hoshi, Takashi; Kawahara, Hiroto; Tominaga, Kenji; Asano, Hiromitsu

    2011-01-01

    LWR (Light Water Reactor) power stations have accumulated many experiences of design, construction and operation. In addition, large-sized reactors have an advantage of economy of scale and 1,000 MWe LWR has therefore become the mainstream reactor in Japan. Meanwhile, introduction of the medium and small-sized LWRs (SMRs) has also been under review in Japan in order to respond to stagnant growth in electricity demand and electricity market liberalization or for investment risk mitigation; however, it has not been realized due to the economic disadvantage of scale. Therefore, JAPC has been developing the concept of SMR (300 MWe - 600 MWe) which is competitive to the large-sized LWR cooperating with Japanese plant makers (Hitachi, Toshiba Corporation and Mitsubishi Heavy Industries), assessing the possibility of realization of SMRs as one of the electric power sources in the future. As the result of the JAPC's study, we developed SMR concepts whose cost and safety are almost equal to large-sized LWR and confirmed technical feasibility of the concept in order to start developing basic design. In this paper, the outline of the SMR concepts and the current development status are presented. Concepts have been developed for two types of SMRs (i.e. BWR and PWR). As for the BWR type, reactor system is simplified by adopting natural circulation core method and CRD falling under gravity in order to downsize the reactor containments. As for the PWR type, the risk of LOCA occurrence is eliminated by unifying the primary system (e.g. incorporating steam generator into reactor). Furthermore, the primary system is simplified by adopting natural circulation core method in operation and containment vessel also become compact for the PWR. As for JAPC's further development of SMRs, key elements of SMR concepts are studied. In addition, the environment surrounding the SMRs has changed in recent years and the one with capacity exceeding 300-600 MWe class or small-sized reactor with

  14. Hydrogen mixing study (HMS) in LWR type containments

    International Nuclear Information System (INIS)

    Travis, J.R.

    1983-01-01

    A numerical technique has been developed for calculating the full three-dimensional time-dependent Navier-Stokes equations with multiple speies transport. The method is a modified form of the Implicit Continuous-fluid Eulerian (ICE) technique to solve the governing equations for low Mach number flows where pressure waves and local variations in compression and expansion are not significant. Large density variations, due to thermal and species concentration gradients, are accounted for without the restrictions of the classical Boussinesq approximation. Calculations of the EPRI/HEDL standard problems verify the feasibility of using this finite-difference technique for analyzing hydrogen mixing within LWR containments

  15. Investigation of valve failure problems in LWR power plants

    Energy Technology Data Exchange (ETDEWEB)

    None

    1980-04-01

    An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

  16. Transmutation of LWR waste actinides in thermal reactors

    International Nuclear Information System (INIS)

    Gorrell, T.C.

    1979-01-01

    Recycle of actinides to a reactor for transmutation to fission products is being considered as a possible means of waste disposal. Actinide transmutation calculations were made for two irradiation options in a thermal (LWR) reactor. The cases considered were: all actinides recycled in regular uranium fuel assemblies, and transuranic actinides recycled in separate mixed oxide (MOX) assemblies. When all actinides were recycled in a uranium lattice, a reduction of 62% in the transuranic inventory was achieved after 10 recycles, compared to the inventory accumulated without recycle. When the transuranics from 2 regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after 5 recycles

  17. Integrity of neutron-absorbing components of LWR fuel systems

    International Nuclear Information System (INIS)

    Bailey, W.J.; Berting, F.M.

    1991-03-01

    A study of the integrity and behavior of neutron-absorbing components of light-water (LWR) fuel systems was performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE). The components studies include control blades (cruciforms) for boiling-water reactors (BWRs) and rod cluster control assemblies for pressurized-water reactors (PWRs). The results of this study can be useful for understanding the degradation of neutron-absorbing components and for waste management planning and repository design. The report includes examples of the types of degradation, damage, or failures that have been encountered. Conclusions and recommendations are listed. 84 refs

  18. Pie technique of LWR fuel cladding fracture toughness test

    International Nuclear Information System (INIS)

    Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu

    2006-01-01

    Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Industries Ltd. (NFI) for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine (EDM), pre-cracking by fatigue tester, sample assembling to the compact tension (CT) shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination (PIE) using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility (WASTEF). (author)

  19. Principles of MONJU maintenance. Characteristic of MONJU maintenance and reflection of LWR maintenance experience to FBR

    International Nuclear Information System (INIS)

    Nakai, Satoru; Nishio, Ryuichi; Uchihashi, Masaya; Kaneko, Yoshihisa; Yamashita, Hironobu; Yamaguchi, Atsunori; Aoki, Takayuki

    2014-01-01

    A sodium cooled fast breeder reactor (FBR) has unique systems and components and different degradation mechanism from light water reactor (LWR) so that need to establish maintenance technology in accordance with its features. The examination of the FBR maintenance technology is carried out in the special committee for considering the maintenance for Monju established in the Japan Society of Maintenology (JSM). As a result of the study such as extraction of Monju maintenance feature, maintenance technology benchmark between Monju and LWR components and survey of LWR maintenance experience, it is clear that principles of maintenance are same as LWR, necessity of LWR maintenance experience reflection and points to be considered in Monju maintenance. The road map to establish a FBR maintenance technology in the technical aspect became clear and it is vital to acquire operation and maintenance experience of the plant to implement this road map, and to establish a fast reactor maintenance. (author)

  20. Dose calculations for severe LWR accident scenarios

    International Nuclear Information System (INIS)

    Margulies, T.S.; Martin, J.A. Jr.

    1984-05-01

    This report presents a set of precalculated doses based on a set of postulated accident releases and intended for use in emergency planning and emergency response. Doses were calculated for the PWR (Pressurized Water Reactor) accident categories of the Reactor Safety Study (WASH-1400) using the CRAC (Calculations of Reactor Accident Consequences) code. Whole body and thyroid doses are presented for a selected set of weather cases. For each weather case these calculations were performed for various times and distances including three different dose pathways - cloud (plume) shine, ground shine and inhalation. During an emergency this information can be useful since it is immediately available for projecting offsite radiological doses based on reactor accident sequence information in the absence of plant measurements of emission rates (source terms). It can be used for emergency drill scenario development as well

  1. Filtered atmospheric venting of LWR containments

    International Nuclear Information System (INIS)

    Hoegberg, L.; Ahlstroem, P.E.; Bachofner, E.; Graeslund, C.; Johansson, K.; Nilsson, L.; Persson, Aa.; Eriksson, B.

    1981-03-01

    The FILTRA project is a cooperative Swedish programme which started in February 1980. It is aimed at investigating the possibility of reducing the risk for a large release of radioactivity, assuming a severe reactor accident. The project has been focused on filtered venting of the reactor containment. The first stage of the project has dealt with two types of severe accident sequences, namely core meltdown as a result of the complete loss of water supplies to the reactor pressure vessel and insufficient cooling of the reactor containment. Some important conclusion are the following. The applicability of computer models used to describe various phenomena in the accident sequence must be scrutinized. The details of the design of the containment are important and must be taken into consideration in a more accurate manner than in previous analyses. A pressure relief area of less than 1 m 2 appears to be adequate. The following principles should guide the technical design of filtered venting systems, namely reduction of the risk for the release of those radioactive substances which could cause long term land contamination, provision for a passive function of the vent filter system during the first 24 hours and achievement of filtering capabilities which make leakages in severe accidents comparable to the leakages of radioactive substances in less severe accidents, which do not necessarily actuate the pressure relief system. Nothing indicates that a system for filtered venting of a BWR containment would have a significant negative effect on the safety within the framework of the design basis. Efforts should be directed towards designing a filtered venting system for a BWR such as Barsebaeck. (authors)

  2. A classification scheme for LWR fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs.

  3. A classification scheme for LWR fuel assemblies

    International Nuclear Information System (INIS)

    Moore, R.S.; Williamson, D.A.; Notz, K.J.

    1988-11-01

    With over 100 light water nuclear reactors operating nationwide, representing designs by four primary vendors, and with reload fuel manufactured by these vendors and additional suppliers, a wide variety of fuel assembly types are in existence. At Oak Ridge National Laboratory, both the Systems Integration Program and the Characteristics Data Base project required a classification scheme for these fuels. This scheme can be applied to other areas and is expected to be of value to many Office of Civilian Radioactive Waste Management programs. To develop the classification scheme, extensive information on the fuel assemblies that have been and are being manufactured by the various nuclear fuel vendors was compiled, reviewed, and evaluated. It was determined that it is possible to characterize assemblies in a systematic manner, using a combination of physical factors. A two-stage scheme was developed consisting of 79 assembly types, which are grouped into 22 assembly classes. The assembly classes are determined by the general design of the reactor cores in which the assemblies are, or were, used. The general BWR and PWR classes are divided differently but both are based on reactor core configuration. 2 refs., 15 tabs

  4. Water chemistry and materials degradation in LWR'S

    International Nuclear Information System (INIS)

    Haenninen, H.; Toerroenen, K.; Aaltonen, P.

    1994-01-01

    Water chemistry plays a major role in corrosion, in erosion corrosion and in activity transport in NPPs; it impacts upon the operational safety of LWRs in two main ways: integrity of pressure boundary materials and activity transport and out-of-core radiation fields. A good control of water chemistry can significantly reduce these problems and improve plant safety, but economic pressures are leading to more rigorous operating conditions: fuel burnups are to be increased, higher efficiencies are to be achieved by running at higher temperatures and plant lifetimes are to be extended. Typical water chemistry specifications used in PWR and BWR plants are presented and the chemistry optimization is discussed. The complex interplay of metallurgical, mechanical and environmental factors in environmental sensitive cracking is shown, with details on studies for carbon steels, stainless steels and nickel base alloys. 20 refs., 8 figs., 4 tabs

  5. PASCAL, Probabilistic Fracture Mechanics Analysis of Structural Components in Aging LWR

    International Nuclear Information System (INIS)

    Shibata, Katsuyuki; Onizawa, Kunio; Li, Yinsheng; Kato, Daisuke

    2005-01-01

    A - Description of program or function: PASCAL (PFM analysis of Structural Components in Aging LWR) is a PFM (Probabilistic Fracture Mechanics) code for evaluating the failure probability of aged pressure components. PASCAL has been developed as a part of the JAERI's research program on aging and structural integrity of LWR components, in order to respond to the increasing need of the probabilistic methodology in the regulation and inspection of nuclear components with the objective to provide a rational tool for the evaluation of the reliability and integrity of structural components. In order to improve the accuracy and reliability of the analysis code, some new fracture mechanics models or computational techniques are introduced considering the recent progress in the state of the art and performance of PC. Thus some new analysis models and original methodologies were introduced in PASCAL such as the elastic-plastic fracture criterion based on R6 method, a new crack extension model of semi-elliptical crack evaluation and so on. Moreover a function to evaluate the effect of embrittlement recovery by annealing of irradiated RPV is also introduced in the code based on the USNRC R.G. 1.162(1996). The code has been verified through various failure analysis results and international PTS round robin analysis ICAS which had been organized by the Principal Working Group 3 of OECD/NEA/CSNI. In order to attain a high usability, PASCAL Ver.1 with GUI provides an exclusive FEM pre-processor Pre-PASCAL for generating the input load transient data, a GUI system for generating the input data for PASCAL main processor of main solver and post-processor for output data. - Pre-PASCAL: Pre-PASCAL is an exclusive 3-D FEM pre-processor for generating the input transient data provided with 3 RPV mesh models and two simple specimen mesh models, i.e. CT and CCP. Almost the same input data format with that of PASCAL main processor is used. Output data of temperature and stress distribution

  6. The dupic fuel cycle synergism between LWR and HWR

    International Nuclear Information System (INIS)

    Lee, J.S.; Yang, M.S.; Park, H.S.; Lee, H.H.; Kim, K.P.; Sullivan, J.D.; Boczar, P.G.; Gadsby, R.D.

    1999-01-01

    The DUPIC fuel cycle can be developed as an alternative to the conventional spent fuel management options of direct disposal or plutonium recycle. Spent LWR fuel can be burned again in a HWR by direct refabrication into CANDU-compatible DUPIC fuel bundles. Such a linkage between LWR and HWR can result in a multitude of synergistic effects, ranging from savings of natural uranium to reductions in the amount of spent fuel to be buried in the earth, for a given amount of nuclear electricity generated. A special feature of the DUPIC fuel cycle is its compliance with the 'Spent Fuel Standard' criteria for diversion resistance, throughout the entire fuel cycle. The DUPIC cycle thus has a very high degree of proliferation resistance. The cost penalty due to this technical factor needs to be considered in balance with the overall benefits of the DUPIC fuel cycle. The DUPIC alternative may be able to make a significant contribution to reducing spent nuclear fuel burial in the geosphere, in a manner similar to the contribution of the nuclear energy alternative in reducing atmospheric pollution from fossil fuel combustion. (author)

  7. Effects of cooling time on a closed LWR fuel cycle

    International Nuclear Information System (INIS)

    Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

    2012-01-01

    In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

  8. Qualification of ARROTTA code for LWR accident analysis

    International Nuclear Information System (INIS)

    Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

    2004-01-01

    This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

  9. Evaluation of LWR fuel rod behavior under operational transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Hiramoto, K.; Maru, A.

    1984-01-01

    To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

  10. OREST, LWR Burnup Simulation Using Program HAMMER and ORIGEN

    International Nuclear Information System (INIS)

    Hesse, Ulrich; Sieberer, Johann

    2006-01-01

    printer-output. 3 - Restrictions on the complexity of the problem: NEA version is limited for 100 loops, 1000 burnup time-steps and 10 post-irradiation steps. GRS recommends the use of LWR fuels based on oxygen and on the main HAMMER isotopes 235-U, 236-U, 238-U, 237-Np, 238-Pu, 239-Pu, 240-Pu, 241-Pu, 242-Pu, 241-Am and 243-Am. Gadolinium entries should be handled with care if singular positions of Gd-rods in real assemblies are found. Other mixture entries at start of calculation should only be impurities. Cladding should be Zr, Al or stainless steel. Special options for handling other materials can be found in the user description. Activation of structure materials is not calculated. Strong heterogeneous assembly problems outside of the input data processor should be pre-calculated by using more-dimensional codes to achieve a neutron spectra equivalent HAMMER lattice (FEC-method). Coolant pressure, coolant temperatures and coolant steam contents are assumed to be constant during burnup. During each program loop neutron spectra and cross sections are assumed to be constant

  11. AFCI : Co-extraction impacts on LWR and fast reactor fuel cycles

    International Nuclear Information System (INIS)

    Taiwo, T. A.; Szakalay, F. J.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2007-01-01

    A systematic investigation of the impact of the co-extraction COEXTM process on reactor performance has been performed. The proliferation implication of the process was also evaluated using the critical mass, radioactivity, decay heat and neutron and gamma source rates and gamma doses as indicators. The use of LWR-spent-uranium-based MOX fuel results in a higher initial plutonium content requirement in an LWR MOX core than if natural uranium based MOX fuel is used (by about 1%); the plutonium for both cases is derived from the spent LWR spent fuel. More transuranics are consequently discharged in the spent fuel of the MOX core. The presence of U-236 in the initial fuel was also found to result in higher content of Np-237 in the spent MOX fuel and less consumption of Pu-238 and Am-241 in the MOX core. The higher quantities of Np-237 (factor of 5), Pu-238 (20%) and Am-241 (14%) decrease the effective repository utilization, relative to the use of natural uranium in the PWR MOX core. Additionally, the minor actinides continue to accumulate in the fuel cycle, even if the U-Pu co-extraction products are continuously recycled in the PWR cores, and thus a solution is required for the minor actinides. The utilization of plutonium derived from LWR spent fuel versus weapons-grade plutonium for the startup core of a 1,000 MWT advanced burner fast reactor (ABR) increases the TRU content by about 4%. Differences are negligible for the equilibrium recycle core. The impact of using reactor spent uranium instead of depleted uranium was found to be relatively smaller in the fast reactor (TRU content difference less than 0.4%). The critical masses of the co-extraction products were found to be higher than that of weapons-grade plutonium and the decay heat and radiation sources of the materials (products) were also found to be generally higher than that of weapons-grade plutonium (WG-Pu) in the transuranics content range of 0.1 to 1.0 in the heavy-metal. The magnitude of the

  12. Development of uranium reduction system for incineration residue generated at LWR nuclear fuel fabrication plants in Japan

    International Nuclear Information System (INIS)

    Sampei, T.; Sato, T.; Suzuki, N.; Kai, H.; Hirata, Y.

    1993-01-01

    The major portion of combustible solid wastes generated at LWR nuclear fuel fabrication plants in Japan is incinerated and stored in a warehouse. The uranium content in the incineration residue is higher compared with other categories of wastes, although only a small amount of incineration residue is generated. Hence, in the future uranium should be removed from incineration residues before they are reduced to a level appropriate for the final disposal. A system for processing the incineration residue for uranium removal has been developed and tested based on the information obtained through laboratory experiments and engineering scale tests

  13. Assessment of management alternatives for LWR wastes. Volume 4. Description of a Belgian scenario for PWR waste

    International Nuclear Information System (INIS)

    Crustin, J.; Glibert, R.

    1993-01-01

    This report deals with the description of a management route for PWR waste relying to a certain extent on Belgian practices in this particular area. This description, which aims at providing input data for subsequent cost evaluation, includes all management steps which are usually implemented for solid, liquid and gaseous wastes from their production up to the interim storage of the final waste products. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  14. Assessment of management alternatives for LWR wastes. Volume 2. Description of a French scenario for PWR waste

    International Nuclear Information System (INIS)

    Saulieu, E. de; Chary, C.

    1993-01-01

    This report deals with the description of a management route for PWR waste relying to a certain extent on French practices in this particular area. This description, which aims at providing input data for subsequent cost evaluation, includes all management steps which are usually implemented for solid, liquid and gaseous wastes from their production up to the interim storage of the final waste products. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  15. Influence of some fabrication parameters and operating conditions on the PCI failure occurrence in LWR fuel rods

    International Nuclear Information System (INIS)

    Bouffioux, P.

    1980-01-01

    In recent LWR designs, the fuel rod failures are induced by a chemically assisted mechanical process, i.e. stress corrosion cracking. The analytical approach towards the analysis of PCI-SCC failures is mainly based on the predictions of the COMETHE code. The failure criteria rely on the concept of a stress threshold together with fission product availability. In the present paper, the use of the COMETHE code to minimize PCI induced clad failure occurrences is illustrated by parametric studies to define acceptable fuel specifications and reactor operating conditions (steady and transient). (author)

  16. BARS - a heterogeneous code for 3D pin-by-pin LWR steady-state and transient calculation

    International Nuclear Information System (INIS)

    Avvakumov, A.V.; Malofeev, V.M.

    2000-01-01

    A 3D pin-by-pin dynamic model for LWR detailed calculation was developed. The model is based on a coupling of the BARS neutronic code with the RELAP5/MOD3.2 thermal hydraulic code. This model is intended to calculate a fuel cycle, a xenon transient, and a wide range of reactivity initiated accidents in a WWER and a PWR. Galanin-Feinberg heterogeneous method was realized in the BARS code. Some results for a validation of the heterogeneous method are presented for reactivity coefficients, a pin-by-pin power distribution, and a fast pulse transient. (Authors)

  17. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  18. Effect of Casmo-5 cross-section data and doppler temperature definitions on LWR reactivity initiated accidents - 166

    International Nuclear Information System (INIS)

    Grandi, G.; Smith, K.; Xu, Z.; Rhodes, J.

    2010-01-01

    During LWR Reactivity Initiated Accidents (RIA), the accurate evaluation of the Doppler reactivity feedback depends on the Doppler coefficient computed by the lattice physics code (e.g. CASMO-5), and on the effective Doppler temperature computed by the transient code (e.g. SIMULATE-3K) using the non-uniform intra-pellet temperature profile. CASMO-5 has many new features compared with its predecessor. Among them, the replacement of the L-library (based primarily on ENDF/B IV data) by the latest available nuclear data (ENDF/B VII.0), and the Monte Carlo based resonance elastic scattering model to overcome deficiencies in NJOY modeling have a significant impact on the fuel temperature coefficient, and hence on LWR RIA. The Doppler temperature effect in thermal reactors is driven by the 238 U absorption. The different effective Doppler temperature definitions, available in the literature, try to capture the considerable self-shielding of the 238 U absorption that occurs in the pellet surface by defining an appropriate fuel temperature to compute cross-sections. In this work, we investigate the effect of the nuclear data generated by CASMO-5 on RIA, as well as the impact of different effective Doppler temperature definitions, including one proposed by the authors. It is concluded: 1) LWR RIA evaluated using CASMO-5 cross section data will be milder because the energy released is ∼10% smaller; 2) the prompt enthalpy rise is barely affected by the choice of the Doppler temperature definition; and 3) the peak fuel enthalpy is affected by the choice of the Doppler temperature definition, the under-prediction of the Doppler reactivity by the 'NEA' Doppler temperature results in a conservative estimate of the peak fuel enthalpy. (authors)

  19. Development of a data bank system for LWR integral experiment

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Aoyagi, Hideo

    1983-01-01

    A data bank system for LWR integral experiment has been developed for the purpose of alleviating various efforts associated with the verification of computer codes. The final aim of this system is such that the imput data for the code to be verified can be easily obtained, and the results of calculation can be obtained in the form of the comparison with measurement. Geometry and material composition as well as measured data are stored in the data bank. This data bank system is composed of four sub-programs; (1) registration program, (2) information retrieval program, (3) maintenance program, and (4) figure representation program. In this report, the structure of this data bank system and how to use the system are explained. An example of the use of this system is also included. (Aoki, K.)

  20. Alternatives for managing post LWR reactor nuclear wastes

    International Nuclear Information System (INIS)

    Platt, A.M.

    1976-01-01

    The two extremes in the LWR fuel cycle are discarding the spent fuel and recycling the U and Pu to the maximum extent possible. The waste volumes from the two alternatives are compared. A preliminary evaluation is made of the technology available for handling wastes from each step of the fuel cycle. The wastes considered are fuel materials, high--level wastes, other liquids, combustible and non-combustible solids, and non--high--level wastes. Evaluation of processing gaseous wastes indicates that technology is available for capture of Kr and I 2 , but further development is needed for T 2 . Technology for interim storage and geological isolation is considered adequate. An outline is given of the steps in the selection of a final storage site

  1. LWR-PV Surveillance Dosimetry Improvement Program review graphics

    International Nuclear Information System (INIS)

    McElroy, W.N.; Gold, R.; Gutherie, G.L.

    1979-10-01

    A primary objective of the multilaboratory program is to prepare an updated and improved set of dosimetry, damage correlation, and the associated reactor analysis ASTM standards for LWR-PV irradiation surveillance programs. Supporting this objective are a series of analytical and experimental validation and calibration studies in Benchmark Neutron Fields, reactor Test Regions, and operating power reactor Surveillance Positions. These studies will establish and certify the precision and accuracy of the measurement and predictive methods which are recommended for use in these standards. Consistent and accurate measurement and data analysis techniques and methods, therefore, will have been developed and validated along with guidelines for required neutron field calculations that are used to (1) correlate changes in material properties with the characteristics of the neutron radiation field and (2) predict pressure vessel steel toughness and embrittlement from power reactor surveillance data

  2. Evaluation of management alternatives for LWR hulls and caps

    International Nuclear Information System (INIS)

    Chaudon, L.; Mehling, O.; Cecille, L.; Thiels, G.; Kowa, S.

    1993-01-01

    Hulls and caps resulting from the reprocessing of LWR spent fuels represent one of the major sources of alpha-bearing solid waste generated during the nuclear fuel cycle. The Commission of the European Communities has undertaken considerable R and D efforts on the development of advanced treatment and conditioning methods for this type of waste. In view of the encouraging results achieved, the Commission launched a theoretical assessment study on cladding waste management. Six practical or potential schemes were identified and elaborated: direct cementation, decontamination prior to cementation, rolling before cementation, rolling followed by embedding in graphite, compaction, and melting in a cold crucible. The economic aspects of each management option were also investigated. This included the assessment of the plant (treatment, conditioning and interim storage), transport and disposal costs. Further consideration will be required to define the best management option for 'cap' wastes. Transport and disposal costs will also require further analysis from an industrial standpoint

  3. Cost optimization of long-cycle LWR operation

    International Nuclear Information System (INIS)

    Handwerk, C.S.; Driscoll, M.J.; McMahon, M.V.; Todreas, N.E.

    1997-01-01

    The continuing emphasis on improvement of plant capacity factor, as a major means to make nuclear energy more cost competitive in the current deregulatory environment, motivates heightened interest in long intra-refueling intervals and high burnup in LWR units. This study examines the economic implications of these trends, to determine the envelope of profitable fuel management tactics. One batch management is found to be significantly more expensive than two-batch management. Parametric studies were carried out varying the most important input parameters. If ultra-high burnup can be achieved, then n = 3 or even n = 4 management may be preferable. For n = 1 or 2, economic performance declines at higher burnups, hence providing no great incentive for moving further in that direction. Values for n > 2 are also attractive because, for a given burnup target, required enrichment decreases as n increases. This study was limited to average batch burnups below 60,000 MWd/MT

  4. Irradiation effects on thermal properties of LWR hydride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  5. Dual-purpose LWR supplying heat for desalination

    International Nuclear Information System (INIS)

    Waplington, G.; Fitcher, H.

    1977-01-01

    A number of desalination processes are at present in various stages of development but distillation is the only serious choice for a large-scale project. The distillation process temperature requirement is low compared with the temperature of steam normally delivered to the turbine in a power generation plant. This gives the possibility for combining the functions of electricity generation with water distillation. The brine heater of the multi-stage flash distillation plant can be supplied with steam after partial expansion through a turbine. Such an arrangement allows the use of a standard nuclear steam supply system and makes fuller use of the energy output than would either single purpose role. The LWR represents a safe, reliable and economic system, and is easily able to provide heat of a quality adequate for the desalination process. (M.S.)

  6. A study on the behavior of defected LWR spent fuel

    International Nuclear Information System (INIS)

    You, Gil Sung; Kim, Eun Ka; Kim, Keon Sik; Suh, Hang Suck; Kim, Seung Jung; Ro, Seung Gy; Park, Chong Mook; Ji, Pyung Gook

    1992-03-01

    To investigate the storage behavior of the defective LWR spent fuel rods, the characteristic changes of fuel and cladding are to be measured and analyzed. In addition, the oxidation study in air on non-irradiated and irradiated U0 2 was performed. No changes were observed in the tested fuel rods after 30 month storage. The Cs-134, 137 released rapidly during the initial 3 months of storage, but remained in constant value after 3 month storage and the release was almost ceased after 30 month storage. The weight gain of non-irradiated U0 2 samples showed a trend of S type curves and the activation energies were 11OKJ/mol above 350 deg C. and 143KJ/mol below 350 deg C. But irradiated U0 2 showed a rapid increase at initial stage of oxidation and a decrease at later stage when compared with the results of non-irradiated U0 2 . (Author)

  7. Development of PIE techniques for irradiated LWR pressure vessel steels

    International Nuclear Information System (INIS)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

    1999-01-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  8. Progress in Development of I2S-LWR Concept

    International Nuclear Information System (INIS)

    Petrovic, Bojan

    2014-01-01

    The paper will present the progress in developing the Integral Inherently Safe Light Water Reactor (12S-LWR) concept. This new concept aims to combine the competitive economics of a large nuclear power plant, with enhanced safety achieved by the integral primary circuit configuration (previously considered only for PWRs with power levels not exceeding several hundred MWc), and with enhanced accident tolerance (to address concerns after the Fukushima Dai-lchi accidents). Several new technologies are being developed to enable this concept, including novel silicide fuel and micro-channel primary heat exchangers. This project is performed by a multi-disciplinary multi-organization team led by Georgia Tech, including academia, a national laboratory, nuclear industry, and a power utility, wit expected participation of the University of Zagreb. (author)

  9. Development of GPT-based optimization algorithm

    International Nuclear Information System (INIS)

    White, J.R.; Chapman, D.M.; Biswas, D.

    1985-01-01

    The University of Lowell and Westinghouse Electric Corporation are involved in a joint effort to evaluate the potential benefits of generalized/depletion perturbation theory (GPT/DTP) methods for a variety of light water reactor (LWR) physics applications. One part of that work has focused on the development of a GPT-based optimization algorithm for the overall design, analysis, and optimization of LWR reload cores. The use of GPT sensitivity data in formulating the fuel management optimization problem is conceptually straightforward; it is the actual execution of the concept that is challenging. Thus, the purpose of this paper is to address some of the major difficulties, to outline our approach to these problems, and to present some illustrative examples of an efficient GTP-based optimization scheme

  10. Diamond-based materials for biomedical applications

    CERN Document Server

    Narayan, Roger

    2013-01-01

    Carbon is light-weight, strong, conductive and able to mimic natural materials within the body, making it ideal for many uses within biomedicine. Consequently a great deal of research and funding is being put into this interesting material with a view to increasing the variety of medical applications for which it is suitable. Diamond-based materials for biomedical applications presents readers with the fundamental principles and novel applications of this versatile material. Part one provides a clear introduction to diamond based materials for medical applications. Functionalization of diamond particles and surfaces is discussed, followed by biotribology and biological behaviour of nanocrystalline diamond coatings, and blood compatibility of diamond-like carbon coatings. Part two then goes on to review biomedical applications of diamond based materials, beginning with nanostructured diamond coatings for orthopaedic applications. Topics explored include ultrananocrystalline diamond for neural and ophthalmologi...

  11. Assessment of management alternatives for LWR wastes. Volume 8. Cost and radiological impact associated with near-surface disposal of reactor waste (Spanish concept)

    International Nuclear Information System (INIS)

    Alamo Berna, S.; Sanchez Delgado, N.

    1993-01-01

    This report deals with the determination of the cost and the radiological impact associated with a near-surface disposal site (Spanish concept) for low and medium-level radioactive waste generated during operation of a 20 GWe nuclear park composed of LWRs for 30 years. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  12. Assessment of management alternatives for LWR wastes. Volume 7. Cost and radiological impact associated with near-surface disposal of reactor waste (French concept)

    International Nuclear Information System (INIS)

    Malherbe, J.

    1993-01-01

    This report deals with the determination of the cost and the radiological impact associated with a near-surface disposal site (French concept) for low and medium-level radioactive waste generated during operation of a 20 GWe nuclear park composed of LWRs for 30 years. This study is part of an overall theoretical exercise aimed at evaluating a selection of management routes for LWR waste based on economical and radiological criteria

  13. A Brief Assessment of North Korea's Capacities for Building an Experimental LWR

    International Nuclear Information System (INIS)

    Lee, Jung Hyu; An, Jin Soo

    2011-01-01

    On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor

  14. A Brief Assessment of North Korea's Capacities for Building an Experimental LWR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Hyu; An, Jin Soo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2011-10-15

    On November 2010, North Korea revealed the construction site of 100 MWt (thermal) experimental LWR in the early stage with a target operation date of 2012. And they claimed that their first LWR construction project is proceeding with strictly domestic talent and resources. Introduction of LWR imposes various technical challenges, even though North Korea has experiences in the construction and management of graphite-moderated and gas-cooled reactor. So, there are doubts about whether they can successfully complete the project in time without any external support. In this paper, to estimate the fate of the LWR construction, we focused on the North Korea's capability to deal with the technical challenges which differ from those of gas-graphite reactor

  15. LWR pressure vessel irradiation surveillance dosimetry. Quarterly progress report, July--September 1978

    Energy Technology Data Exchange (ETDEWEB)

    Guthrie, G L; McElroy, W N; Lippincott, E P; Gold, R

    1978-12-01

    Program objectives and progress to date by the national laboratories in LWR pressure vessel irradiation surveillance dosimetry are summarized. Participants in the program include: Rockwell International, Hanford Engineering Development Laboratory, National Bureau of Standards, and Oak Ridge National Laboratory.

  16. Irradiated test fuel shipment plan for the LWR MOX fuel irradiation test project

    International Nuclear Information System (INIS)

    Shappert, L.B.; Dickerson, L.S.; Ludwig, S.B.

    1998-01-01

    This document outlines the responsibilities of DOE, DOE contractors, the commercial carrier, and other organizations participating in a shipping campaign of irradiated test specimen capsules containing mixed-oxide (MOX) fuel from the Idaho National Engineering and Environmental Laboratory (INEEL) to the Oak Ridge National Laboratory (ORNL). The shipments described here will be conducted according to applicable regulations of the US Department of Transportation (DOT), US Nuclear Regulatory Commission (NRC), and all applicable DOE Orders. This Irradiated Test Fuel Shipment Plan for the LWR MOX Fuel Irradiation Test Project addresses the shipments of a small number of irradiated test specimen capsules and has been reviewed and agreed to by INEEL and ORNL (as participants in the shipment campaign). Minor refinements to data entries in this plan, such as actual shipment dates, exact quantities and characteristics of materials to be shipped, and final approved shipment routing, will be communicated between the shipper, receiver, and carrier, as needed, using faxes, e-mail, official shipping papers, or other backup documents (e.g., shipment safety evaluations). Any major changes in responsibilities or data beyond refinements of dates and quantities of material will be prepared as additional revisions to this document and will undergo a full review and approval cycle

  17. Probabilistic safety assessment of LWR containment systems performance. Report of principal working group n.5 on risk assessment

    International Nuclear Information System (INIS)

    Holloway, N.J.; Harper, F.T.; Bellard, S.W.

    1992-01-01

    This report reviews current approaches to PSA of LWR containment systems performance. It is based on a variety of recent PSA reports which deal with Level-2 PSA. The report is a summary of recent state-of-the-art containment analysis and is intended to assist analysts in their selection of the most appropriate methods of extending Level-1 plant safety evaluations into Level-2 assessments of the containment performance. The document is primarily concerned with the performance of the containment as an engineered system rather than with the source terms consequent upon its failure. It is addressed mainly to the performance of large dry PWR containments, with a secondary emphasis on other containment types. After explaining the purposes of these analyses, a survey of LWR containment analysis options is presented: direct approaches using containment event tree construction, indirect approaches based on previous PSAs, alternative and novel approaches. The selection process is then described, followed by conclusions on their suitability for various cases: accident management, research prioritization, identifying design weaknesses, specific issue resolution, modelling physical reality, etc.

  18. Developing context sensitive BIM based applications

    NARCIS (Netherlands)

    Hartmann, Timo; Underwood, J.; Isikdag, U.

    2010-01-01

    Current Building Information Model (BIM) based applications do not integrate well with the varying and frequently changing work processes of Architectural, Engineering, and Construction (AEC) professionals. One cause for this problem is that traditionally software developers apply software design

  19. Cloud Based Applications and Platforms (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Brodt-Giles, D.

    2014-05-15

    Presentation to the Cloud Computing East 2014 Conference, where we are highlighting our cloud computing strategy, describing the platforms on the cloud (including Smartgrid.gov), and defining our process for implementing cloud based applications.

  20. Designing Android Based Augmented Reality Location-Based Service Application

    Directory of Open Access Journals (Sweden)

    Alim Hardiansyah

    2018-01-01

    Full Text Available Android is an operating system for Linux based smartphone. Android provides an open platform for the developers to create their own application. The most developed and used application now is location based application. This application gives personalization service for mobile device user and is customized to their location. Location based service also gives an opportunity for the developers to develop and increase the value of service. One of the technologies that could be combined with location based application is augmented reality. Augmented reality combines the virtual world with the real one. By the assistance of augmented reality, our surrounding environment could interact in digital form. Information of objects and environment surround us could be added to the augmented reality system and presented. Based on the background, the writers tried to implement those technologies on now rapidly developing android application as a final project to achieve bachelor degree in Department of Informatics Engineering, Faculty of Information Technology and Visual Communication, Al Kamal Science and Technology Institute. This application could be functioned to locate school by using location based service technology with the assistance of navigational applications such as waze and google maps, in form of live direction process through the smartphone

  1. Behavior of LWR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

    1977-01-01

    In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

  2. A fracture mechanics approach for estimating fatigue crack initiation in carbon and low-alloy steels in LWR coolant environments

    International Nuclear Information System (INIS)

    Park, H. B.; Chopra, O. K.

    2000-01-01

    A fracture mechanics approach for elastic-plastic materials has been used to evaluate the effects of light water reactor (LWR) coolant environments on the fatigue lives of carbon and low-alloy steels. The fatigue life of such steel, defined as the number of cycles required to form an engineering-size crack, i.e., 3-mm deep, is considered to be composed of the growth of (a) microstructurally small cracks and (b) mechanically small cracks. The growth of the latter was characterized in terms of ΔJ and crack growth rate (da/dN) data in air and LWR environments; in water, the growth rates from long crack tests had to be decreased to match the rates from fatigue S-N data. The growth of microstructurally small cracks was expressed by a modified Hobson relationship in air and by a slip dissolution/oxidation model in water. The crack length for transition from a microstructurally small crack to a mechanically small crack was based on studies on small crack growth. The estimated fatigue S-N curves show good agreement with the experimental data for these steels in air and water environments. At low strain amplitudes, the predicted lives in water can be significantly lower than the experimental values

  3. Impact of implicit effects on uncertainties and sensitivities of the Doppler coefficient of a LWR pin cell

    Science.gov (United States)

    Hursin, Mathieu; Leray, Olivier; Perret, Gregory; Pautz, Andreas; Bostelmann, Friederike; Aures, Alexander; Zwermann, Winfried

    2017-09-01

    In the present work, PSI and GRS sensitivity analysis (SA) and uncertainty quantification (UQ) methods, SHARK-X and XSUSA respectively, are compared for reactivity coefficient calculation; for reference the results of the TSUNAMI and SAMPLER modules of the SCALE code package are also provided. The main objective of paper is to assess the impact of the implicit effect, e.g., considering the effect of cross section perturbation on the self-shielding calculation, on the Doppler coefficient SA and UQ. Analyses are done for a Light Water Reactor (LWR) pin cell based on Phase I of the UAM LWR benchmark. The negligence of implicit effects in XSUSA and TSUNAMI leads to deviations of a few percent between the sensitivity profiles compared to SAMPLER and TSUNAMI (incl. implicit effects) except for 238U elastic scattering. The implicit effect is much larger for the SHARK-X calculations because of its coarser energy group structure between 10 eV and 10 keV compared to the applied SCALE libraries. It is concluded that the influence of the implicit effect strongly depends on the energy mesh of the nuclear data library of the neutron transport solver involved in the UQ calculations and may be magnified by the response considered.

  4. Successful case-based reasoning applications 2

    CERN Document Server

    Jain, Lakhmi

    2014-01-01

    Case-based reasoning paradigms offer automatic reasoning capabilities which are useful for the implementation of human like machines in a limited sense. This research book is the second volume in a series devoted to presenting Case-based reasoning (CBR) applications. The first volume, published in 2010, testified the flexibility of CBR, and its applicability in all those fields where experiential knowledge is available. This second volume further witnesses the heterogeneity of the domains in which CBR can be exploited, but also reveals some common directions that are clearly emerging in recent years. This book will prove useful to the application engineers, scientists, professors and students who wish to develop successful case-based reasoning applications.

  5. Fracture toughness evaluation of select advanced replacement alloys for LWR core internals

    Energy Technology Data Exchange (ETDEWEB)

    Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chen, Xiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Fracture toughness is one of the key engineering properties required for core internal materials. Together with other properties, which are being examined such as high-temperature steam oxidation resistance, radiation hardening, and irradiation-assisted stress corrosion cracking resistance, the alloys will be down-selected for neutron irradiation study and comprehensive post-irradiation examinations. According to the candidate alloys selected under the ARRM program, ductile fracture toughness of eight alloys was evaluated at room temperature and the LWR-relevant temperatures. The tested alloys include two ferritic alloys (Grade 92 and an oxide-dispersion-strengthened alloy 14YWT), two austenitic stainless steels (316L and 310), four Ni-base superalloys (718A, 725, 690, and X750). Alloy 316L and X750 are included as reference alloys for low- and high-strength alloys, respectively. Compact tension specimens in 0.25T and 0.2T were machined from the alloys in the T-L and R-L orientations according to the product forms of the alloys. This report summarizes the final results of the specimens tested and analyzed per ASTM Standard E1820. Unlike the

  6. Democratic People's Republic of Korea LWR project status

    International Nuclear Information System (INIS)

    Mulligan, J.B.

    1996-01-01

    In October 1994, at Geneva, the United States and the Democratic People's Republic of Korea (DPRK) signed an Agreed Framework as a first step toward resolving international concerns about nuclear activities in the DPRK. This Agreement, when implemented, will ultimately lead to the complete dismantlement of those aspects of the DPRK's nuclear program, including reprocessing-related facilities, that have undermined the viability of the international nuclear non-proliferation regime and the stability of the Asia-Pacific region. The essence of the Agreement is that the DPRK will take near-term action to cease the activities of concern and permit some International Atomic Energy Agency (IAEA) verification inspection. In the future, it will dismantle its production reactors and accept full-scope IAWA safeguards. In return, the United Stated agreed to lead an international effort to supply the DPRK with light-water reactors which are less of proliferation concern than are graphite-moderated production reactors. Until the first LWR is in operation the DPRK will receive shipments of heavy oil to replace the energy lost by shutting down the production reactors

  7. CCGT + LWR = the power plant of the future?

    International Nuclear Information System (INIS)

    Tsiklauri, G.

    1997-01-01

    The thermal efficiency of LWR type reactors can be increased making use of the Tsikl-Durst cycle, where the gas turbine is combined with the nuclear reactor using a steam mixer. The principle of this combined cycle is outlined. It is envisaged that the overall thermal efficiency of the power plant can be increased to 41 - 44%. The total output would be two to three times higher. With advanced light-water reactors (ABWR, AP-600) and advanced gas turbines in combination with the one-way steam generator as developed by Solar Turbines Inc., producing steam at 650 degC to 750 degC, it is feasible to attain a total thermal efficiency of 55%. The combination of two kinds of fuel (nuclear fuel and natural gas) improves operating flexibility of the cycle in various regimes so as to respond to natural gas prices and electricity demands. The gas turbine adds to the nuclear power plant an independent source of power, so that standby dieselgenerators are no more necessary. (P.A.). 1 tab., 2 figs

  8. Qualification of the neutronic evolution of LWR fuels in MELUSINE

    International Nuclear Information System (INIS)

    Beretz, D.; Garcin, J.; Ducros, G.; Vanhumbeeck, D.; Chaucheprat, P.

    1984-09-01

    MELUSINE, a swimming pool type reactor, in Grenoble, for research and technological irradiations is well fitted to the neutronic evolution qualification of the LWR fuel. Thus, with an adjustment of the lattice pitch, representative neutron spectrum locations are available. The re-leading management and the regulation mode flexibility of MELUSINE lead to reproductible neutronic parameters configurations without restricting the reactor to this purpose only. Under these conditions, simple calculations can be carried out for interpretation, without taking into account the whole core. An instrumentation by Self Power Neutron Detectors (collectrons) gives on-line information on the fluxes at the periphery of the device. When required by the neutronicians, experimental pins can be unloaded during the irradiation process and scanned on a gammametry bench immersed in the reactor-pool itself, before their isotopic composition analysis. Thus, within the framework of neutronic evolution qualification, are studied fuel pins for advanced assemblies for the light water reactors or their derivatives, with large advantages over irradiations in power reactors [fr

  9. Convergence studies of deterministic methods for LWR explicit reflector methodology

    International Nuclear Information System (INIS)

    Canepa, S.; Hursin, M.; Ferroukhi, H.; Pautz, A.

    2013-01-01

    The standard approach in modem 3-D core simulators, employed either for steady-state or transient simulations, is to use Albedo coefficients or explicit reflectors at the core axial and radial boundaries. In the latter approach, few-group homogenized nuclear data are a priori produced with lattice transport codes using 2-D reflector models. Recently, the explicit reflector methodology of the deterministic CASMO-4/SIMULATE-3 code system was identified to potentially constitute one of the main sources of errors for core analyses of the Swiss operating LWRs, which are all belonging to GII design. Considering that some of the new GIII designs will rely on very different reflector concepts, a review and assessment of the reflector methodology for various LWR designs appeared as relevant. Therefore, the purpose of this paper is to first recall the concepts of the explicit reflector modelling approach as employed by CASMO/SIMULATE. Then, for selected reflector configurations representative of both GII and GUI designs, a benchmarking of the few-group nuclear data produced with the deterministic lattice code CASMO-4 and its successor CASMO-5, is conducted. On this basis, a convergence study with regards to geometrical requirements when using deterministic methods with 2-D homogenous models is conducted and the effect on the downstream 3-D core analysis accuracy is evaluated for a typical GII deflector design in order to assess the results against available plant measurements. (authors)

  10. MCNP analysis of the nine-cell LWR gadolinium benchmark

    International Nuclear Information System (INIS)

    Arkuszewski, J.J.

    1988-01-01

    The Monte Carlo results for a 9-cell fragment of the light water reactor square lattice with a central gadolinium-loaded pin are presented. The calculations are performed with the code MCNP-3A and the ENDF-B/5 library and compared with the results obtained from the BOXER code system and the JEF-1 library. The objective of this exercise is to study the feasibility of BOXER for the analysis of a Gd-loaded LWR lattice in the broader framework of GAP International Benchmark Analysis. A comparison of results indicates that, apart from unavoidable discrepancies originating from different data evaluations, the BOXER code overestimates the multiplication factor by 1.4 % and underestimates the power release in a Gd cell by 4.66 %. It is hoped that further similar studies with use of the JEF-1 library for both BOXER and MCNP will help to isolate and explain these discrepancies in a cleaner way. (author) 4 refs., 9 figs., 10 tabs

  11. Validation of KENOREST with LWR-PROTEUS phase II samples

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, M.; Kilger, R.; Pautz, A.; Zwermann, W. [GRS, Garching (Germany); Grimm, P.; Vasiliev, A.; Ferroukhi, H. [Paul Scherrer Institut, Villigen (Switzerland)

    2012-11-01

    In order to broaden the validation basis of the reactivity and nuclide inventory code KENOREST two samples of the LWR-PROTEUS phase II program have been calculated and compared to the experimental results. In general most nuclides are reproduced very well and agree within about ten percent with the experiment. Some already known problems, the overprediction of metallic fission products and the underprediction of the higher curium isotopes, have been confirmed. One of the largest uncertainties in the calculation was the burnup of the samples due to differences between a core simulation of the fuel vendor and the burnup determined from the measured values of the burnup indicator Nd-148. Two different models taking into account the environment for a peripheral fuel rod have been studied. The more detailed model included the three direct neighbor fuel assemblies depleted along with the fuel rod of interest. The influence on the results has been found to be very small. Compared to the uncertainties from the burnup, this effect can be considered negligible. The reason for the low influence was basically that the spectrum did not get considerably harder with increasing burnup beyond about 20GWd/tHM. Since the sample reached burnups far beyond that value, an effect could not be seen. In the near future an update of the used libraries is planned and it will be very interesting to study the effect on the results, especially for Curium. (orig.)

  12. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Heuser, Brent [Univ. of Illinois, Urbana-Champaign, IL (United States); Stubbins, James [Univ. of Illinois, Urbana-Champaign, IL (United States); Kozlowski, Tomasz [Univ. of Illinois, Urbana-Champaign, IL (United States); Uddin, Rizwan [Univ. of Illinois, Urbana-Champaign, IL (United States); Trinkle, Dallas [Univ. of Illinois, Urbana-Champaign, IL (United States); Downar, Thoms [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); ang, Yong [Univ. of Florida, Gainesville, FL (United States); Phillpot, Simon [Univ. of Florida, Gainesville, FL (United States); Sabharwall, piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-25

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  13. Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel

    International Nuclear Information System (INIS)

    Heuser, Brent; Stubbins, James; Kozlowski, Tomasz; Uddin, Rizwan; Trinkle, Dallas; Downar, Thoms; Was, Gary; Ang, Yong; Phillpot, Simon; Sabharwall, Piyush

    2017-01-01

    The DOE NEUP sponsored IRP on accident tolerant fuel (ATF) entitled Engineered Zircaloy Cladding Modifications for Improved Accident Tolerance of LWR Nuclear Fuel involved three academic institutions, Idaho National Laboratory (INL), and ATI Materials (ATI). Detailed descriptions of the work at the University of Illinois (UIUC, prime), the University of Florida (UF), the University of Michigan (UMich), and INL are included in this document as separate sections. This summary provides a synopsis of the work performed across the IRP team. Two ATF solution pathways were initially proposed, coatings on monolithic Zr-based LWR cladding material and selfhealing modifications of Zr-based alloys. The coating pathway was extensively investigated, both experimentally and in computations. Experimental activities related to ATF coatings were centered at UIUC, UF, and UMich and involved coating development and testing, and ion irradiation. Neutronic and thermal hydraulic aspects of ATF coatings were the focus of computational work at UIUC and UMich, while materials science aspects were the focus of computational work at UF and INL. ATI provided monolithic Zircaloy 2 and 4 material and a binary Zr-Y alloy material. The selfhealing pathway was investigated with advanced computations only. Beryllium was identified as a valid self-healing additive early in this work. However, all attempts to fabricate a Zr-Be alloy failed. Several avenues of fabrication were explored. ATI ultimately declined our fabrication request over health concerns associated with Be (we note that Be was not part of the original work scope and the ATI SOW). Likewise, Ames Laboratory declined our fabrication request, citing known litigation dating to the 1980s and 1990s involving the U.S. Federal government and U.S. National Laboratory employees involving the use of Be. Materion (formerly, Brush Wellman) also declined our fabrication request, citing the difficulty in working with a highly reactive Zr and Be

  14. Rate Theory Modeling and Simulations of Silicide Fuel at LWR Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, Bei [Argonne National Lab. (ANL), Argonne, IL (United States); Mei, Zhigang [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, Gerard [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-10

    Uranium silicide (U3Si2) fuel has higher thermal conductivity and higher uranium density, making it a promising candidate for the accident-tolerant fuel (ATF) used in light water reactors (LWRs). However, previous studies on the fuel performance of U3Si2, including both experimental and computational approaches, have been focusing on the irradiation conditions in research reactors, which usually involve low operation temperatures and high fuel burnups. Thus, it is important to examine the fuel performance of U3Si2 at typical LWR conditions so as to evaluate the feasibility of replacing conventional uranium dioxide fuel with this silicide fuel material. As in-reactor irradiation experiments involve significant time and financial cost, it is appropriate to utilize modeling tools to estimate the behavior of U3Si2 in LWRs based on all those available research reactor experimental references and state-of-the-art density functional theory (DFT) calculation capabilities at the early development stage. Hence, in this report, a comprehensive investigation of the fission gas swelling behavior of U3Si2 at LWR conditions is introduced. The modeling efforts mentioned in this report was based on the rate theory (RT) model of fission gas bubble evolution that has been successfully applied for a variety of fuel materials at devious reactor conditions. Both existing experimental data and DFT-calculated results were used for the optimization of the parameters adopted by the RT model. Meanwhile, the fuel-cladding interaction was captured by the coupling of the RT model with simplified mechanical correlations. Therefore, the swelling behavior of U3Si2 fuel and its consequent interaction with cladding in LWRs was predicted by the rate theory modeling, providing valuable information for the development of U3Si2 fuel as an accident

  15. Performance of a transmutation advanced device for sustainable energy application

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, C.; Rosales, J.; Garcia, L. [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Perez-Navarro, A.; Escriva, A. [Universidad Politecnica de Valencia, Valencia (Spain). Inst. de Ingenieria Energetica; Abanades, A. [Universidad Politecnica de Madrid (Spain). Grupo de Modelizacion de Sistemas Termoenergeticos

    2009-07-01

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

  16. Performance of a transmutation advanced device for sustainable energy application

    International Nuclear Information System (INIS)

    Garcia, C.; Rosales, J.; Garcia, L.; Perez-Navarro, A.; Escriva, A.; Abanades, A.

    2009-01-01

    Preliminary studies have been performed to design a device for nuclear waste transmutation and hydrogen generation based on a gas cooled pebble bed accelerator driven system, TADSEA (transmutation advanced device for sustainable energy application). In previous studies we have addressed the viability of an ADS Transmutation device that uses as fuel wastes from the existing LWR power plants, encapsulated in graphite in the form of pebble beds, being cooled by helium which enables high temperatures, in the order of 1200 K, to facilitate hydrogen generation from water either by high temperature electrolysis or by thermo chemical cycles. To design this device several configurations were studied, including several reactors thickness, to achieve the desired parameters, the transmutation of nuclear waste and the production of 100 MW. of thermal power. In this paper we are presenting new studies performed on deep burn in-core fuel management strategy for LWR waste. We analyze the fuel cycle on TADSEA device based on driver and transmutation fuel that were proposed for the General Atomic design of a gas turbine-modular helium reactor. We compare the transmutation results of the three fuel management strategies, using driven and transmutation, and standard LWR spend fuel, and present several parameters that describe the neutron performance of TADSEA nuclear core as the fuel and moderator temperature reactivity coefficients and transmutation chain. (author)

  17. New Cork-Based Materials and Applications

    Directory of Open Access Journals (Sweden)

    Luís Gil

    2015-02-01

    Full Text Available This review work is an update of a previous work reporting the new cork based materials and new applications of cork based materials. Cork is a material which has been used for multiple applications. The most known uses of cork are in stoppers (natural and agglomerated cork for alcoholic beverages, classic floor covering with composite cork tiles (made by the binding of cork particles with different binders, and thermal/acoustic/vibration insulation with expanded corkboard in buildings and some other industrial fields. Many recent developments have been made leading to new cork based materials. Most of these newly developed cork materials are not yet on the market, but they represent new possibilities for engineers, architects, designers and other professionals which must be known and considered, potentially leading to their industrialization. This paper is a review covering the last five years of innovative cork materials and applications also mentioning previous work not reported before.

  18. New Cork-Based Materials and Applications

    Science.gov (United States)

    Gil, Luís

    2015-01-01

    This review work is an update of a previous work reporting the new cork based materials and new applications of cork based materials. Cork is a material which has been used for multiple applications. The most known uses of cork are in stoppers (natural and agglomerated cork) for alcoholic beverages, classic floor covering with composite cork tiles (made by the binding of cork particles with different binders), and thermal/acoustic/vibration insulation with expanded corkboard in buildings and some other industrial fields. Many recent developments have been made leading to new cork based materials. Most of these newly developed cork materials are not yet on the market, but they represent new possibilities for engineers, architects, designers and other professionals which must be known and considered, potentially leading to their industrialization. This paper is a review covering the last five years of innovative cork materials and applications also mentioning previous work not reported before. PMID:28787962

  19. New Cork-Based Materials and Applications.

    Science.gov (United States)

    Gil, Luís

    2015-02-10

    This review work is an update of a previous work reporting the new cork based materials and new applications of cork based materials. Cork is a material which has been used for multiple applications. The most known uses of cork are in stoppers (natural and agglomerated cork) for alcoholic beverages, classic floor covering with composite cork tiles (made by the binding of cork particles with different binders), and thermal/acoustic/vibration insulation with expanded corkboard in buildings and some other industrial fields. Many recent developments have been made leading to new cork based materials. Most of these newly developed cork materials are not yet on the market, but they represent new possibilities for engineers, architects, designers and other professionals which must be known and considered, potentially leading to their industrialization. This paper is a review covering the last five years of innovative cork materials and applications also mentioning previous work not reported before.

  20. Cellulose-Based Nanomaterials for Energy Applications.

    Science.gov (United States)

    Wang, Xudong; Yao, Chunhua; Wang, Fei; Li, Zhaodong

    2017-11-01

    Cellulose is the most abundant natural polymer on earth, providing a sustainable green resource that is renewable, degradable, biocompatible, and cost effective. Recently, nanocellulose-based mesoporous structures, flexible thin films, fibers, and networks are increasingly developed and used in photovoltaic devices, energy storage systems, mechanical energy harvesters, and catalysts components, showing tremendous materials science value and application potential in many energy-related fields. In this Review, the most recent advancements of processing, integration, and application of cellulose nanomaterials in the areas of solar energy harvesting, energy storage, and mechanical energy harvesting are reviewed. For solar energy harvesting, promising applications of cellulose-based nanostructures for both solar cells and photoelectrochemical electrodes development are reviewed, and their morphology-related merits are discussed. For energy storage, the discussion is primarily focused on the applications of cellulose-based nanomaterials in lithium-ion batteries, including electrodes (e.g., active materials, binders, and structural support), electrolytes, and separators. Applications of cellulose nanomaterials in supercapacitors are also reviewed briefly. For mechanical energy harvesting, the most recent technology evolution in cellulose-based triboelectric nanogenerators is reviewed, from fundamental property tuning to practical implementations. At last, the future research potential and opportunities of cellulose nanomaterials as a new energy material are discussed. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  1. ZZ SAIL, Albedo Scattering Data Library for 3-D Monte-Carlo Radiation Transport in LWR Pressure Vessel

    International Nuclear Information System (INIS)

    1982-01-01

    1 - Description of problem or function: Format: SAIL format; Number of groups: 23 neutron / 17 gamma-ray; Nuclides: Type 04 Concrete and Low Carbon Steel (A533B). Origin: Science Applications, Inc (SAI); Weighting spectrum: yes. SAIL is a library of albedo scattering data to be used in three-dimensional Monte Carlo codes to solve radiation transport problems specific to the reactor pressure vessel cavity region of a LWR. The library contains data for Type 04 Concrete and Low Carbon Steel (A533B). 2 - Method of solution: The calculation of the albedo data was perform- ed with a version of the discrete ordinates transport code DOT which treats the transport of neutrons, secondary gamma-rays and gamma- rays in one dimension, while maintaining the complete two-dimension- al treatment of the angular dependence

  2. Engineering applications using CAD based application programming interface

    Directory of Open Access Journals (Sweden)

    Tzotzis Anastasios

    2017-01-01

    Full Text Available Automating the design process of a product or a system can provide engineers and designers with many benefits. As such, repeatable tasks that are time consuming can be handled automatically and with minimal human attention. This is achieved by using the API (Application Programmable Interface of CAD systems in order to create macros or even develop software applications. The present paper deals with an application that has been developed with the API of a general purposes CAD system. This application automates the design process of a standard pneumatic double acting cylinder based on the appropriate inserted parameters (ISO 15552.The design process begins with the creation of a series of components developed as solids, and extends to the extraction of basic attributes and properties from the complete mechanical assembly. Finally, the assembled models and the extracted data can be used to further study the design of the pneumatic double acting cylinder. It is expected in the future to expand the features of the presented application in order to automate the design process of other related mechanical systems.

  3. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    Energy Technology Data Exchange (ETDEWEB)

    Lori Braase

    2013-01-01

    The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), in collaboration with the nuclear industry, has been conducting research and development (R&D) activities on advanced Light Water Reactor (LWR) fuels for the last few years. The emphasis for these activities was on improving the fuel performance in terms of increased burnup for waste minimization and increased power density for power upgrades, as well as collaborating with industry on fuel reliability. After the events at the Fukushima Nuclear Power Plant in Japan in March 2011, enhancing the accident tolerance of LWRs became a topic of serious discussion. In the Consolidated Appropriations Act, 2012, Conference Report 112-75, the U.S. Congress directed DOE-NE to: • Give “priority to developing enhanced fuels and cladding for light water reactors to improve safety in the event of accidents in the reactor or spent fuel pools.” • Give “special technical emphasis and funding priority…to activities aimed at the development and near-term qualification of meltdown-resistant, accident-tolerant nuclear fuels that would enhance the safety of present and future generations of light water reactors.” • Report “to the Committee, within 90 days of enactment of this act, on its plan for development of meltdown-resistant fuels leading to reactor testing and utilization by 2020.” Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system currently used by the nuclear industry, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, and operational transients, as well as design-basis and beyond design-basis events. The overall draft strategy for development and demonstration is comprised of three phases: Feasibility Assessment and Down-selection; Development and Qualification; and

  4. Flooding of a large, passive, pressure-tube LWR

    Energy Technology Data Exchange (ETDEWEB)

    Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1995-09-01

    A reactor concept has been developed which can survive LOCA without scram and without replenishing primary coolant inventory. The proposed concept is a pressure tube type reactor similar to CANDU reactors, but differing in three key aspects: (1) a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles, (2) the heavy water coolant in the pressure tubes is replaced by light water, and (3) the calandria tank contains a low pressure gas instead of heavy water moderator. The gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents, it allows passive calandria flooding. This paper describes the thermal hydraulic characteristics of the gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube LWR concept. The flooding of the top row of fuel channels must be accomplished fast enough so that none of the critical components of the fuel channel exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

  5. Analysis of SCC initiation/propagation behavior of stainless steels in LWR environments

    International Nuclear Information System (INIS)

    Saito, Koichi; Kuniya, Jiro

    1999-01-01

    This paper presents a method to analyze initiation and propagation behavior of stress corrosion cracking (SCC) of stainless steels on the basis of a new prediction algorithm in which the initiation period and propagation period of SCC under irradiation conditions are considered from a practical viewpoint. The prediction algorithm is based on three ideas: (1) threshold neutron fluence of radiation-enhanced SCC (RESCC), (2) equivalent critical crack depth, and (3) threshold stress intensity factor for SCC (K ISCC ). SCC initiation/propagation behavior in light water reactor (LWR) environments is analyzed by incorporating model equations on irradiation hardening, irradiation-enhanced electrochemical potentiokinetic reactivation (EPR) and irradiation stress relaxation that are phenomena peculiar to neutron irradiation. The analytical method is applied to predict crack growth behavior of a semi-elliptical surface crack in a flat plane that has an arbitrary residual stress profile; specimens are sensitized type 304 stainless steels which had been subjected to neutron irradiation in high temperature water. SCC growth behavior of a semi-elliptical surface crack was greatly dependent on the distribution of residual stress in a flat plane. When residual stress at the surface of the flat plane was relatively small, the method predicted SCC propagation did not take place. (author)

  6. ORCENT-2, Full Load Steam Turbine Cycle Thermodynamics for LWR Power Plant

    International Nuclear Information System (INIS)

    Fuller, L.C.

    1979-01-01

    1 - Description of problem or function: ORCENT-2 performs heat and mass balance calculations at valves-wide-open design conditions, maximum guaranteed rating conditions, and an approximation of part-load conditions for steam turbine cycles supplied with throttle steam, characteristic of contemporary light-water reactors. The program handles both condensing and back-pressure turbine exhaust arrangements. Turbine performance calculations are based on the General Electric Company method for 1800-rpm large steam turbine- generators operating with light-water-cooled nuclear reactors. Output includes all information normally shown on a turbine-cycle heat balance diagram. 2 - Method of solution: The turbine performance calculations follow the procedures outlined in General Electric report GET-6020. ORCENT-2 utilizes the 1967 American Society of Mechanical Engineers (ASME) formulations and procedures for calculating the properties of steam, adapted for ORNL use by D.W. Altom. 3 - Restrictions on the complexity of the problem: Maxima of: 12 feed-water heaters, 5 moisture removal stages in the low-pressure turbine section. ORCENT-2 is limited to 1800-rpm tandem-compound turbine-generators with single- or double-flow high pressure sections and one, two, or three double-flow low-pressure turbine sections. Steam supply for LWR cycles should be between 900 and 1100 psia and slightly wet to 100 degrees F of initial superheat. Generator rating should be greater than 100 MVA

  7. Estimation of irradiation-induced material damage measure of FCM fuel in LWR core

    International Nuclear Information System (INIS)

    Lee, Kyung-Hoon; Lee, Chungchan; Park, Sang-Yoon; Cho, Jin-Young; Chang, Jonghwa; Lee, Won Jae

    2014-01-01

    An irradiation-induced material damage measure on tri-isotropic (TRISO) multi-coating layers of fully ceramic micro-encapsulated (FCM) fuel to replace conventional uranium dioxide (UO 2 ) fuel for existing light water reactors (LWRs) has been estimated using a displacement per atom (DPA) cross section for a FCM fuel performance analysis. The DPA cross sections in 47 and 190 energy groups for both silicon carbide (SiC) and graphite are generated based on the molecular dynamics simulation by SRIM/TRIM. For the selected FCM fuel assembly design with FeCrAl cladding, a core depletion analysis was carried out using the DeCART2D/MASTER code system with the prepared DPA cross sections to evaluate the irradiation effect in the Korean OPR-1000. The DPA of the SiC and IPyC coating layers is estimated by comparing the discharge burnup obtained from the MASTER calculation with the burnup-dependent DPA for each coating layer calculated using DeCART2D. The results show that low uranium loading and hardened neutron spectrum compared to that of high temperature gas-cooled reactor (HTGR) result in high discharge burnup and high fast neutron fluence. In conclusion, it can be seen that the irradiation-induced material damage measure is noticeably increased under LWR operating conditions compared to HTGRs. (author)

  8. NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2016-01-01

    Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

  9. Hydrogen removal from LWR containments by catalytic-coated thermal insulation elements (THINCAT)

    International Nuclear Information System (INIS)

    Fischer, K.; Broeckerhoff, P.; Ahlers, G.; Gustavsson, V.; Herranz, L.; Polo, J.; Dominguez, T.; Royl, P.

    2003-01-01

    In the THINCAT project, an alternative concept for hydrogen mitigation in a light water reactor (LWR) containment is being developed. Based on catalytic coated thermal insulation elements of the main coolant loop components, it could be considered either as an alternative to backfitting passive autocatalytic recombiner devices, or as a reinforcement of their preventive effect. The present paper summarises the results achieved at about project mid-term. Potential advantages of catalytic thermal insulation studied in the project are:-reduced risk of unintended ignition,;-no work space obstruction in the containment,;-no need for seismic qualification of additional equipment,;-improved start-up behaviour of recombination reaction. Efforts to develop a suitable catalytic layer resulted in the identification of a coating procedure that ensures high chemical reactivity and mechanical stability. Test samples for use in forthcoming experiments with this coating were produced. Models to predict the catalytic rates were developed, validated and applied in a safety analysis study. Results show that an overall hydrogen concentration reduction can be achieved which is comparable to the reduction obtained using conventional recombiners. Existing experimental information supports the argument of a reduced ignition risk

  10. I2S-LWR Activation Analysis of Heat Exchangers Using Hybrid Shielding Methodology with SCALE6.1

    International Nuclear Information System (INIS)

    Matijevic, M.; Pevec, D.; Jecmenica, R.

    2016-01-01

    The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typical for Generation III+ small modular reactors. The main safety feature is based on integral primary circuit configuration, bringing together compact design of the reactor core with 121 fuel assembly (FA), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on silicide fuel is selected to achieve a high thermal power which is extracted with PHEs placed in the annual region between the barrel and the vessel. The complex and integrated design of I2S-LWR leads to activation of integral components, mainly made from stainless steel, so accurate and precise Monte Carlo (MC) simulations are needed to quantify potential dose rates to personnel during routine maintenance operation. This shielding problem is therefore very challenging one, posing a non-trivial neutron flux solution in a phase space. This paper presents the performance of the hybrid shielding methodologies CADIS/FW-CADIS implemented in the MAVRIC sequence of the SCALE6.1 code package. The main objective was to develop a detailed MC shielding model of the I2S-LWR reactor along with effective variance reduction (VR) parameters and to calculate neutron fluence rates inside PHEs. Such results are then utilized to find neutron activation rate distribution via 60Co generation inside of a stack of microchannel heat exchangers (MCHX), which will be periodically withdrawn for the maintenance. 59Co impurities are the main cause of (n,gamma) radiative gamma dose to personnel via neutron activation since 60Co has half-life of 5.27 years and is emitting high energy gamma rays (1.17 MeV and 1.33 MeV). The developed MC model was successfully used to find converged fluxes inside all 8 stacks of

  11. Vision based systems for UAV applications

    CERN Document Server

    Kuś, Zygmunt

    2013-01-01

    This monograph is motivated by a significant number of vision based algorithms for Unmanned Aerial Vehicles (UAV) that were developed during research and development projects. Vision information is utilized in various applications like visual surveillance, aim systems, recognition systems, collision-avoidance systems and navigation. This book presents practical applications, examples and recent challenges in these mentioned application fields. The aim of the book is to create a valuable source of information for researchers and constructors of solutions utilizing vision from UAV. Scientists, researchers and graduate students involved in computer vision, image processing, data fusion, control algorithms, mechanics, data mining, navigation and IC can find many valuable, useful and practical suggestions and solutions. The latest challenges for vision based systems are also presented.

  12. Duality based optical flow algorithms with applications

    DEFF Research Database (Denmark)

    Rakêt, Lars Lau

    We consider the popular TV-L1 optical flow formulation, and the so-called duality based algorithm for minimizing the TV-L1 energy. The original formulation is extended to allow for vector valued images, and minimization results are given. In addition we consider different definitions of total...... variation regularization, and related formulations of the optical flow problem that may be used with a duality based algorithm. We present a highly optimized algorithmic setup to estimate optical flows, and give five novel applications. The first application is registration of medical images, where X......-ray images of different hands, taken using different imaging devices are registered using a TV-L1 optical flow algorithm. We propose to regularize the input images, using sparsity enhancing regularization of the image gradient to improve registration results. The second application is registration of 2D...

  13. Investigation of Nuclide Importance to Functional Requirements Related to Transport and Long-Term Storage of LWR Spent Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, B.L.

    1995-01-01

    100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with LWR spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC).

  14. Technical development on burn-up credit for spent LWR fuels

    International Nuclear Information System (INIS)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  15. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  16. Effects of Listening While Reading (LWR on Swahili Reading Fluency and Comprehension

    Directory of Open Access Journals (Sweden)

    Filipo Lubua

    2016-10-01

    Full Text Available A number of studies have examined the contribution of technology in teaching such languages as English, French, and Spanish, among many others. Contrarily, most LCTL’s, have received very little attention. This study investigates if listening while reading (LWR may expedite Swahili reading fluency and comprehension. The study employed the iBook Author tool to create weekly mediated and interactive reading texts, with comprehension exercises, which were eventually used to collect descriptive and qualitative data from four Elementary Swahili students. Participants participated in a seven week reading program, which provided them with some kind of directed self-learning, and met with the instructor for at least 30 minutes every week for observation and more reading activities. The teacher recorded their reading scores, and a number of themes on how LWR influenced reading fluency and comprehension are discussed here. It shows that participants have a positive attitude towards LWR and they suggest it for all the reading classes.

  17. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  18. EUV lithography for 22nm half pitch and beyond: exploring resolution, LWR, and sensitivity tradeoffs

    Science.gov (United States)

    Putna, E. Steve; Younkin, Todd R.; Leeson, Michael; Caudillo, Roman; Bacuita, Terence; Shah, Uday; Chandhok, Manish

    2011-04-01

    The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 22nm half pitch node and beyond. According to recent assessments made at the 2010 EUVL Symposium, the readiness of EUV materials remains one of the top risk items for EUV adoption. The main development issue regarding EUV resists has been how to simultaneously achieve high resolution, high sensitivity, and low line width roughness (LWR). This paper describes our strategy, the current status of EUV materials, and the integrated post-development LWR reduction efforts made at Intel Corporation. Data collected utilizing Intel's Micro- Exposure Tool (MET) is presented in order to examine the feasibility of establishing a resist process that simultaneously exhibits <=22nm half-pitch (HP) L/S resolution at <=11.3mJ/cm2 with <=3nm LWR.

  19. Web-based applications for virtual laboratories

    NARCIS (Netherlands)

    Bier, H.H.

    2011-01-01

    Web-based applications for academic education facilitate, usually, exchange of multimedia files, while design-oriented domains such as architectural and urban design require additional support in collaborative real-time drafting and modeling. In this context, multi-user interactive interfaces

  20. Value-based medicine: concepts and application

    Directory of Open Access Journals (Sweden)

    Jong-Myon Bae

    2015-03-01

    Full Text Available Global healthcare in the 21st century is characterized by evidence-based medicine (EBM, patient-centered care, and cost effectiveness. EBM involves clinical decisions being made by integrating patient preference with medical treatment evidence and physician experiences. The Center for Value-Based Medicine suggested value-based medicine (VBM as the practice of medicine based upon the patient-perceived value conferred by an intervention. VBM starts with the best evidence-based data and converts it to patient value-based data, so that it allows clinicians to deliver higher quality patient care than EBM alone. The final goals of VBM are improving quality of healthcare and using healthcare resources efficiently. This paper introduces the concepts and application of VBM and suggests some strategies for promoting related research.

  1. Value-based medicine: concepts and application

    Science.gov (United States)

    Bae, Jong-Myon

    2015-01-01

    Global healthcare in the 21st century is characterized by evidence-based medicine (EBM), patient-centered care, and cost effectiveness. EBM involves clinical decisions being made by integrating patient preference with medical treatment evidence and physician experiences. The Center for Value-Based Medicine suggested value-based medicine (VBM) as the practice of medicine based upon the patient-perceived value conferred by an intervention. VBM starts with the best evidence-based data and converts it to patient value-based data, so that it allows clinicians to deliver higher quality patient care than EBM alone. The final goals of VBM are improving quality of healthcare and using healthcare resources efficiently. This paper introduces the concepts and application of VBM and suggests some strategies for promoting related research. PMID:25773441

  2. Value-based medicine: concepts and application.

    Science.gov (United States)

    Bae, Jong-Myon

    2015-01-01

    Global healthcare in the 21st century is characterized by evidence-based medicine (EBM), patient-centered care, and cost effectiveness. EBM involves clinical decisions being made by integrating patient preference with medical treatment evidence and physician experiences. The Center for Value-Based Medicine suggested value-based medicine (VBM) as the practice of medicine based upon the patient-perceived value conferred by an intervention. VBM starts with the best evidence-based data and converts it to patient value-based data, so that it allows clinicians to deliver higher quality patient care than EBM alone. The final goals of VBM are improving quality of healthcare and using healthcare resources efficiently. This paper introduces the concepts and application of VBM and suggests some strategies for promoting related research.

  3. Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 1. Summary: alternatives for the back of the LWR fuel cycle types and properties of LWR fuel cycle wastes projections of waste quantities; selected glossary

    International Nuclear Information System (INIS)

    1976-05-01

    Volume I of the five-volume report contains executive and technical summaries of the entire report, background information of the LWR fuel cycle alternatives, descriptions of waste types, and projections of waste quantities. Overview characterizations of alternative LWR fuel cycle modes are also included

  4. Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

    Energy Technology Data Exchange (ETDEWEB)

    Miller, W.J.

    1980-05-01

    This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.

  5. Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

    International Nuclear Information System (INIS)

    Miller, W.J.

    1980-05-01

    This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan

  6. Comparison of scale/triton and helios burnup calculations for high burnup LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Tittelbach, S.; Mispagel, T.; Phlippen, P.W. [WTI Wissenschaftlich-Technische Ingenieurberatung GmbH, Juelich (Germany)

    2009-07-01

    The presented analyses provide information about the suitability of the lattice burnup code HELIOS and the recently developed code SCALE/TRITON for the prediction of isotopic compositions of high burnup LWR fuel. The accurate prediction of the isotopic inventory of high burnt spent fuel is a prerequisite for safety analyses in and outside of the reactor core, safe loading of spent fuel into storage casks, design of next generation spent fuel casks and for any consideration of burnup credit. Depletion analyses are performed with both burnup codes for PWR and BWR fuel samples which were irradiated far beyond 50 GWd/t within the LWR-PROTEUS Phase II project. (orig.)

  7. Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

    2014-04-01

    Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

  8. Use of phenomena identification and ranking (PIRT) process in research related to design certification of the AP600 advanced passive light water reactor (LWR)

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Eltawila, F.

    1996-01-01

    The AP600 LWR is a new advanced passive design that has been submitted to the USNRC for design certification. Within the certification process the USNRC will perform selected system thermal hydraulic response audit studies to help confirm parts of the vendor's safety analysis submittal. Because of certain innovative design features of the safety systems, new experimental data and related advances in the system thermal hydraulic analysis computer code are being developed by the USNRC. The PIRT process is being used to focus the experimental and analytical work to obtain a sufficient and cost effective research effort. The objective of this paper is to describe the application and most significant results of the PIRT process, including several innovative features needed in the application to accommodate the short design certification schedule. The short design certification schedule has required that many aspects of the USNRC experimental and analytical research be performed in parallel, rather than in series as was normal for currently operating LWRS. This has required development and use of management techniques that focus and integrate the various diverse parts of the research. The original PIRTs were based on inexact knowledge of an evolving reactor design, and concentrated on the new passive features of the design. Subsequently, the PIRTs have evolved in two more stages as the design became more firm and experimental and analytical data became available. A fourth and final stage is planned and in progress to complete the PIRT development. The PIRTs existing at the end of each development stage have been used to guide the experimental program, scaling analyses and code development supporting the audit studies

  9. Design Application Early Childhood Education Based Mobile

    Directory of Open Access Journals (Sweden)

    Annah

    2016-02-01

    Full Text Available This study aims to make learning media for Early Childhood Education in the form especially its mobile applications for Android-based smartphones. In the process of teaching and learning for Early Childhood Education is still often found constraints experienced teaching force is limited props so that learners are less eager to learn. In addition, parents also have difficulty returning to guide or teach the learning materials at home because it has no alternative instructional media. In compiling this research report author uses the Android-based Mobile Devices Applications created using the Java programming language through the Eclipse editor. Based on the results of the research, concluded that these applications can be applied in the latest version of the Android platform to its current platform version of Jellybean. Application of Learning can be used as an alternative way of learning for Early Childhood Education so as to overcome the lack of props in institutions of Early Childhood Education, can be used to be taught at home, and provide new teaching methods to early childhood so that a form of learning that is obtained is not the monotony of one form of learning how.

  10. Reputation based security model for android applications

    OpenAIRE

    Tesfay, Welderufael Berhane; Booth, Todd; Andersson, Karl

    2012-01-01

    The market for smart phones has been booming in the past few years. There are now over 400,000 applications on the Android market. Over 10 billion Android applications have been downloaded from the Android market. Due to the Android popularity, there are now a large number of malicious vendors targeting the platform. Many honest end users are being successfully hacked on a regular basis. In this work, a cloud based reputation security model has been proposed as a solution which greatly mitiga...

  11. Security Assessment of Web Based Distributed Applications

    Directory of Open Access Journals (Sweden)

    Catalin BOJA

    2010-01-01

    Full Text Available This paper presents an overview about the evaluation of risks and vulnerabilities in a web based distributed application by emphasizing aspects concerning the process of security assessment with regards to the audit field. In the audit process, an important activity is dedicated to the measurement of the characteristics taken into consideration for evaluation. From this point of view, the quality of the audit process depends on the quality of assessment methods and techniques. By doing a review of the fields involved in the research process, the approach wants to reflect the main concerns that address the web based distributed applications using exploratory research techniques. The results show that many are the aspects which must carefully be worked with, across a distributed system and they can be revealed by doing a depth introspective analyze upon the information flow and internal processes that are part of the system. This paper reveals the limitations of a non-existing unified security risk assessment model that could prevent such risks and vulnerabilities debated. Based on such standardize models, secure web based distributed applications can be easily audited and many vulnerabilities which can appear due to the lack of access to information can be avoided.

  12. TRU transmutation using ThO2-UO2 and fully ceramic micro-encapsulated fuels in LWR fuel assemblies

    International Nuclear Information System (INIS)

    Bae, Gonghoon; Hong, Sergi

    2012-01-01

    The objective of this work is to design new LWR fuel assemblies which are able to efficiently destroy TRU (transuranics) nuclide without degradation of safety aspects by using ThO 2 -UO 2 fuel pins and FCM (Fully Ceramic Micro-encapsulated) fuel pins containing TRU fuel particles. Thorium was mixed to UO 2 in order to reduce the generation of plutonium nuclides and to save the uranium resources in the UO 2 pins. Additionally, the use of thorium contributes to the extension of the fuel cycle length. All calculations were performed by using DeCART (Deterministic Core Analysis based on Ray Tracing) code. The results show that the new concept of fuel assembly has the TRU destruction rates of ∼40% and ∼25% per 1200 EFPD (Effective Full Power Day) over the TRU FCM pins and the overall fuel assembly, respectively, without degradation of FTC and MTC

  13. Experience gained in the current LWR that influence the design and operation of the LWR advanced from the viewpoint of safety analysis

    International Nuclear Information System (INIS)

    Barrera, J.; Corisco, M.; Riverola, J.

    2010-01-01

    Since the construction of the first light water reactors (LWR) safety analysis has played a very important role in the operation and its evolution to come up with designs that are currently operating. With new tools available, this role will see increased allowing more efficient operation with security assessments in real time, and a more efficient designs both in terms of fuel efficiency and from the security of the plant during operation.

  14. LWR safety research in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Seipel, H.G.

    1977-01-01

    The paper gives a review of the German LWR safety research programme. It describes how the programme was initiated and informs on its goals, development andpractical realization, and indicates how it is bound up with international collaboration. The contribution so far made by the programme to an enhancement of the understanding of major safety problems and to the improvement of safety technology is demonstrated by means of a few selected examples. Experiments relating to loss-of--coolant accidents have deepened our understanding of the heat transfer in the reactor core during blowdown as well as during the flooding phase. Investigations of the dynamic effects going on in dry full pressure containments and pressure suppression systems, following a loss-of--coolant accident, have indicated that existing computer models cannot satisfactorily predict all relevant physical phenomena. Yet, the experimental results obtained constitute a sufficient basis for safe containment design. Research work on core meltdown accidents has identified the particular importance of the type of concrete used for the containment structures and its foundation. If basaltic concrete is used, a substantial fission product release to the environment is extremely unlikely even in the case of a core meltdown accident. At least, it would take place much later than was previously assumed. Resrach on the safety of pressurized components has been concentrated on the problem of cracks in the heat-affected zone of welds. New methods were developed for the detection and analysis of the acceptability of microcrack fields. Additional investigations of specimens and components to increase the understanding of the long-term behaviour of components with microcracks are envisaged in the frame of a new major project on ''component safety''. Considerable progress has been made in the development of methods for automatic remote-control volumetric testing of reactor pressure vessels using ultrasonic techniques

  15. Pulse oximeter based mobile biotelemetry application.

    Science.gov (United States)

    Işik, Ali Hakan; Güler, Inan

    2012-01-01

    Quality and features of tele-homecare are improved by information and communication technologies. In this context, a pulse oximeter-based mobile biotelemetry application is developed. With this application, patients can measure own oxygen saturation and heart rate through Bluetooth pulse oximeter at home. Bluetooth virtual serial port protocol is used to send the test results from pulse oximeter to the smart phone. These data are converted into XML type and transmitted to remote web server database via smart phone. In transmission of data, GPRS, WLAN or 3G can be used. The rule based algorithm is used in the decision making process. By default, the threshold value of oxygen saturation is 80; the heart rate threshold values are 40 and 150 respectively. If the patient's heart rate is out of the threshold values or the oxygen saturation is below the threshold value, an emergency SMS is sent to the doctor. By this way, the directing of an ambulance to the patient can be performed by doctor. The doctor for different patients can change these threshold values. The conversion of the result of the evaluated data to SMS XML template is done on the web server. Another important component of the application is web-based monitoring of pulse oximeter data. The web page provides access to of all patient data, so the doctors can follow their patients and send e-mail related to the evaluation of the disease. In addition, patients can follow own data on this page. Eight patients have become part of the procedure. It is believed that developed application will facilitate pulse oximeter-based measurement from anywhere and at anytime.

  16. Transient stratification modelling of a corium pool in a LWR vessel lower head

    International Nuclear Information System (INIS)

    Le Tellier, R.; Saas, L.; Bajard, S.

    2015-01-01

    Highlights: • A kinetic stratification model is proposed for the simulation of the in-vessel corium behaviour during a LWR severe accident. • The different associated “modes” of vessel failure by thermal focusing effect are highlighted and discussed. • A sensitivity study for a 1650 MWe GenIII PWR is presented with this model in order to illustrate the associated R&D issues. - Abstract: In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters

  17. Advanced in-situ characterisation of corrosion properties of LWR fuel cladding materials

    International Nuclear Information System (INIS)

    Arilahti, E.; Bojinov, M.; Beverskog, B.

    1999-01-01

    The trend towards higher fuel burnups imposes a demand for better corrosion and hydriding resistance of cladding materials. Development of new and improved cladding materials is a long process. There is a lack of fast and reliable in-situ techniques to investigate zirconium alloys in simulated or in-core LWR coolant conditions. This paper describes a Thin Layer Electrode (TLE) arrangement suitable for in-situ characterization of oxide films formed on fuel cladding materials. This arrangement enables us to carry out: Versatile Thin Layer Electrochemical measurements, including: (i) Thin Layer Electrochemical impedance Spectroscopic (TLEIS) measurements to characterize the oxidation kinetics and mechanisms of metals and the properties of their oxide films in aqueous environments. These measurements can also be performed in low conductivity electrolytes. (ii) Thin-Layer Wall-Jet (TLWJ) measurements, which give the possibility to detect soluble reaction products and to evaluate the influence of novel water chemistry additions on their release. Solid Contact measurements: (i) Contact Electric Resistance (CER) measurements to investigate the electronic properties of surface films on the basis of d.c. resistance measurements. (i) Contact Electric impedance (CEI) measurements to study the electronic properties of surface films using a.c. perturbation. All the above listed measurements can be performed using one single measurement device developed at VTT. This device can be conveniently inserted into an autoclave. Its geometry is currently being optimized in cooperation with the OECD Halden Reactor Project. In addition, the applicability of the device for in-core measurements has been investigated in a joint feasibility study performed by VTT and JRC Petten. Results of some autoclave studies of the effect of LiOH concentration on the stability of fuel cladding oxide films are presented in this paper. (author)

  18. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    International Nuclear Information System (INIS)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment

  19. Reduction of nuclear waste burden from LWR by deployment of the SCNES

    International Nuclear Information System (INIS)

    Arie, Kazuo; Watanabe, Junko; Mori, Kenji; Kubota, Kenichi; Kawashima, Masatoshi; Nakayama, Yoshiyuki; Nakazono, Ryuichi; Kuroda, Yuji; Fujiie, Yoichi

    2009-01-01

    Current efforts for enhancing capabilities for energy generation by LWR systems are efficient against the global warming crisis. In parallel to those movements, early realization of the SCNES concept can be the most viable solution to reduce nuclear waste burden produced by the current energy production system. (author)

  20. Air quality impacts due to construction of LWR waste management facilities

    International Nuclear Information System (INIS)

    1977-06-01

    Air quality impacts of construction activities and induced housing growth as a result of construction activities were evaluated for four possible facilities in the LWR fuel cycle: a fuel reprocessing facility, fuel storage facility, fuel fabrication plant, and a nuclear power plant. Since the fuel reprocessing facility would require the largest labor force, the impacts of construction of that facility were evaluated in detail

  1. Computer program of iodine removal in the LWR containment vessel under LOCA conditions, MIRA-PB

    International Nuclear Information System (INIS)

    Nishio, Gunji; Tanaka, Mitsugu; Tamura, Tomohiko.

    1978-03-01

    LWR plants have a containment system for reactor safety consisting of spray and air cleaning filter. R.L.Ritzman of Battele Columbus Lab. developed computer code MIRAP/MIRAB for predicting iodine removal by containment system for PWR and BWR; which has some problem, however. The computer code MIRA-PB prepared by the authors is a modification of MIRAP/MIRAB. (auth.)

  2. PIE of nuclear grade SiC/SiC flexural coupons irradiated to 10 dpa at LWR temperature

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    Silicon carbide fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230–340°C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials are chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC)-coated Hi-NicalonTM Type-S (HNS), TyrannoTM SA3 (SA3), and SCS-UltraTM (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young’s modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young’s moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

  3. Fission product releases at severe LWR accident conditions: ORNL/CEA measurements versus calculations

    Energy Technology Data Exchange (ETDEWEB)

    Andre, B.; Ducros, G.; Leveque, J.P. [CEA Centre d`Etudes de Grenoble, 38 (France). Dept. de Thermohydraulique et de Physique; Osborne, M.F.; Lorenz, R.A. [Oak Ridge National Lab., TN (United States); Maro, D. [CEA Centre d`Etudes de Fontenay-aux-Roses, 92 (France). Dept. de Protection de l`Environnement et des Installations

    1995-12-31

    Experimental programs in the United States and France have followed similar paths in supplying much of the data needed to analyze severe accidents. Both the HI/VI program, conducted at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the U. S. Nuclear Regulatory Commission (NRC), and the HEVA/VERCORS program, supported by IPSN-Commissariat a l`Energie Atomique (CEA) and carried out at the Centre d`Etudes Nucleaires de Grenoble, have studied fission product release from light water reactor (LWR) fuel samples during test sequences representative of severe accidents. Recognizing that more accurate data, i.e., a better defined source term, could reduce the safety margins included in the rather conservative source terms originating from WASH-1400, the primary objective of these programs has been to improve the data base concerning fission product release and behavior at high temperatures. To facilitate the comparison, a model based on fission product diffusion mechanisms that was developed at ORNL and adapted with CEA experimental data is proposed. This CEA model is compared with the ORNL experimental data in a blind test. The two experimental programs used similar techniques in out-of-pile studies. Highly irradiated fuel samples were heated in radiofrequency induction furnaces to very high temperatures (up to 2700 K at ORNL and 2750 K at CEA) in oxidizing (H{sub 2}O), reducing (H{sub 2}) or mixed (H{sub 2}O+H{sub 2}) environments. The experimental parameters, which were chosen from calculated accident scenarios, did not duplicate specific accidents, but rather emphasized careful control of test conditions to facilitate extrapolation of the results to a wide variety of accident situations. This paper presents a broad and consistent database from ORNL and CEA release results obtained independently since the early 1980`S. A comparison of CORSOR and CORSOR Booth calculations, currently used in safety analysis, and the experimental results is presented and

  4. Triarylborane-Based Materials for OLED Applications

    Directory of Open Access Journals (Sweden)

    Gulsen Turkoglu

    2017-09-01

    Full Text Available Multidisciplinary research on organic fluorescent molecules has been attracting great interest owing to their potential applications in biomedical and material sciences. In recent years, electron deficient systems have been increasingly incorporated into fluorescent materials. Triarylboranes with the empty p orbital of their boron centres are electron deficient and can be used as strong electron acceptors in conjugated organic fluorescent materials. Moreover, their applications in optoelectronic devices, energy harvesting materials and anion sensing, due to their natural Lewis acidity and remarkable solid-state fluorescence properties, have also been investigated. Furthermore, fluorescent triarylborane-based materials have been commonly utilized as emitters and electron transporters in organic light emitting diode (OLED applications. In this review, triarylborane-based small molecules and polymers will be surveyed, covering their structure-property relationships, intramolecular charge transfer properties and solid-state fluorescence quantum yields as functional emissive materials in OLEDs. Also, the importance of the boron atom in triarylborane compounds is emphasized to address the key issues of both fluorescent emitters and their host materials for the construction of high-performance OLEDs.

  5. Characterisation of high-burnup LWR fuel rods through gamma tomography

    International Nuclear Information System (INIS)

    Caruso, S.

    2007-01-01

    Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235 U enrichment of the fresh UO 2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO 2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

  6. Safety requirements applicable to the SMART design

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Kim, Wee Kyong; Kim, Hho Jung

    1999-01-01

    The 330 MW thermal power of integral reactor, named SMART (System integrated Modular Advanced ReacTor), is under development at KAERI for seawater desalination application and electricity generation. The final product of nuclear desalination plant (NDP) is electricity and fresh water. Thus, in addition to the protection of the public around the plant facility from the possible release of radioactive materials, the fresh water should be prevented from radioactivity contamination. In this study, to ensure the safety of SMART reactor in the early stage of design development, the safety requirements applicable to the SMART design were investigated, based on the current regulatory requirements for the existing NPPs and the advanced light water reactor (LWR) designs. The interface requirements related to the desalination facility were also investigated, based on the recent IAEA research activities pertaining to the NDP. As a result, it was found that the current regulatory requirements and guidance for the existing NPPs and advanced LWR designs are applicable to the SMART design and its safety evaluation. However, the safety requirements related to the SMART-specific design and the desalination plant are needed to develop in the future to assure the safety of the SMART reactor

  7. Assessment of management alternatives for LWR wastes. Volume 5. Assessment of the radiological impact to the public resulting from discharges of radioactive effluents

    International Nuclear Information System (INIS)

    Centner, B.

    1993-01-01

    This report deals with the assessment of the radiological impact to the public resulting from discharges of radioactive effluents (liquid and gaseous) in connection with the implementation of the Belgian scenario for the management of PWR waste. Both individual and collective doses have been estimated for a critical group of the population living around the nuclear power plants concerned. This study is part of an overall theoretical exercise aimed at evaluating a selection of management wastes for LWR waste based on economical and radiological criteria

  8. Graphene-based fibers for supercapacitor applications

    International Nuclear Information System (INIS)

    Chen, Lianlian; Zhao, Yang; Chen, Nan; Qu, Liangti; Liu, Yu

    2016-01-01

    Energy conversion and storage devices play an important role in industry and society with the rapid growth of energy consumption. Supercapacitors are very attractive due to their superior power density, fast charge/discharge rates and long cycle lifetime. Graphene fiber (GF), a fascinating material, has drawn considerable attention and shown great potential as an active material in the field of supercapacitors owing to its unique and tunable nanostructure, high electrical conductivity, excellent mechanical flexibility, light weight, and ease of functionalization. This review focuses on the recent significant advances in the fabrication and application of graphene-based fiber as electrode material in supercapacitors. The synthetic strategies and application in the supercapacitor are presented, accompanied with the summary and outlook for the future development of GFs. (topical review)

  9. PBG based terahertz antenna for aerospace applications

    CERN Document Server

    Choudhury, Balamati; Jha, Rakesh Mohan

    2016-01-01

    This book focuses on high-gain antennas in the terahertz spectrum and their optimization. The terahertz spectrum is an unallocated EM spectrum, which is being explored for a number of applications, especially to meet increasing demands of high data rates for wireless space communications. Space communication systems using the terahertz spectrum can resolve the problems of limited bandwidth of present wireless communications without radio-frequency interference. This book describes design of such high-gain antennas and their performance enhancement using photonic band gap (PBG) substrates. Further, optimization of antenna models using evolutionary algorithm based computational engine has been included. The optimized high-performance compact antenna may be used for various wireless applications, such as inter-orbital communications and on-vehicle satellite communications.

  10. Orthogonally Based Digital Content Management Applicable to Projects-bases

    Directory of Open Access Journals (Sweden)

    Daniel MILODIN

    2009-01-01

    Full Text Available There is defined the concept of digital content. The requirements of an efficient management of the digital content are established. There are listed the quality characteristics of digital content. Orthogonality indicators of digital content are built up. They are meant to measure the image, the sound as well as the text orthogonality as well. Projects-base concept is introduced. There is presented the model of structuring the content in order to maximize orthogonality via a convergent iterative process. The model is instantiated for the digital content of a projects-base. It is introduced the application used to test the model. The paper ends with conclusions.

  11. Application of the GIF PR and PP methodology to a commercial fast reactor system for a preliminary analysis of PR scenarios

    International Nuclear Information System (INIS)

    Rossi, Fabiana

    2015-01-01

    The Generation IV International Forum (GIF) Proliferation Resistance and Physical Protection (PR and PP) Working Group has developed a methodology for the PR and PP evaluation of the next generation Nuclear Energy Systems (NESs). Following the methodology proposed by the working group, applicable to assessing the proliferation resistance of an NES and its individual elements, the main objective of this work is to apply the methodology to show an example of how its results could be used by designers to improve the PR of the system. In this study, the reactor site of a hypothetical and commercial sodium‑cooled fast neutron nuclear reactor system (SFR) was used as the target NES for the application of the methodology. The design of this SFR is based on the layout of the Japanese Sodium Fast Reactor (JSFR) with a safeguards design based on the safeguards approach of the Japanese prototype fast breeder reactor Monju. The methodology is applied to all the PR scenarios described in the methodology: diversion, misuse and breakout. The methodology was first applied to the SFR to check if this system meets the target of PR as described in the GIF goal; secondly, a comparison between the SFR and a light water reactor (LWR) with an open fuel cycle was performed to evaluate if and how it would be possible to improve the PR and PP of the SFR. The LWR layout is based on the European Pressurized Water Reactor. The comparison was implemented according to the following example development target: achieving proliferation resistance to material diversion similar or superior to domestic and international advanced LWR. Three main actions were performed: implement the evaluation methodology based on its assumptions; characterize the PR and PP for the nuclear energy system applying the methodology to the SFR; and identify recommendations for system designers through comparing the SFR with the LWR.

  12. Ultrasonically identified cap seal for LWR fuel bundles

    International Nuclear Information System (INIS)

    Buergers, W.; Dal Cero, J.; Crutzen, S.

    1981-01-01

    This paper aims to provide a general review of techniques available for surveillance and for sealing, marking or otherwise identifying material in such a way that its recognition and guarantee of integrity are unequivocally assured. The problem of obtaining such assurance has been the subject of work at Ispra and elsewhere. Some discussion of the problems, the work performed and possible solution are given. In addition, techniques which, although not yet in routine use, would be suitable for such application are described. Using industrial ultrasonic apparatus, the signal obtained by scanning the seals was very satisfactory, as was shown by the evaluation studies. The general method is based on reflection due to the great difference in acoustical impedance existing between the matrix and the inclusions

  13. Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

    International Nuclear Information System (INIS)

    Jasiulevicius, Audrius

    2003-01-01

    This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

  14. Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Jasiulevicius, Audrius

    2003-07-01

    This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

  15. Voxel-Based LIDAR Analysis and Applications

    Science.gov (United States)

    Hagstrom, Shea T.

    One of the greatest recent changes in the field of remote sensing is the addition of high-quality Light Detection and Ranging (LIDAR) instruments. In particular, the past few decades have been greatly beneficial to these systems because of increases in data collection speed and accuracy, as well as a reduction in the costs of components. These improvements allow modern airborne instruments to resolve sub-meter details, making them ideal for a wide variety of applications. Because LIDAR uses active illumination to capture 3D information, its output is fundamentally different from other modalities. Despite this difference, LIDAR datasets are often processed using methods appropriate for 2D images and that do not take advantage of its primary virtue of 3-dimensional data. It is this problem we explore by using volumetric voxel modeling. Voxel-based analysis has been used in many applications, especially medical imaging, but rarely in traditional remote sensing. In part this is because the memory requirements are substantial when handling large areas, but with modern computing and storage this is no longer a significant impediment. Our reason for using voxels to model scenes from LIDAR data is that there are several advantages over standard triangle-based models, including better handling of overlapping surfaces and complex shapes. We show how incorporating system position information from early in the LIDAR point cloud generation process allows radiometrically-correct transmission and other novel voxel properties to be recovered. This voxelization technique is validated on simulated data using the Digital Imaging and Remote Sensing Image Generation (DIRSIG) software, a first-principles based ray-tracer developed at the Rochester Institute of Technology. Voxel-based modeling of LIDAR can be useful on its own, but we believe its primary advantage is when applied to problems where simpler surface-based 3D models conflict with the requirement of realistic geometry. To

  16. Application of an integrated PC-based neutronics code system to criticality safety

    International Nuclear Information System (INIS)

    Briggs, J.B.; Nigg, D.W.

    1991-01-01

    An integrated system of neutronics and radiation transport software suitable for operation in an IBM PC-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past four years. Four modules within the system are particularly useful for criticality safety applications. Using the neutronics portion of the integrated code system, effective neutron multiplication values (k eff values) have been calculated for a variety of benchmark critical experiments for metal systems (Plutonium and Uranium), Aqueous Systems (Plutonium and Uranium) and LWR fuel rod arrays. A description of the codes and methods used in the analysis and the results of the benchmark critical experiments are presented in this paper. In general, excellent agreement was found between calculated and experimental results. (Author)

  17. Effect of long-term storage of LWR spent fuel on Pu-thermal fuel cycle

    International Nuclear Information System (INIS)

    Kurosawa, Masayoshi; Naito, Yoshitaka; Suyama, Kenya; Itahara, Kuniyuki; Suzuki, Katsuo; Hamada, Koji

    1998-01-01

    According to the Long-term Program for Research, Development and Utilization of Nuclear Energy (June, 1994) in Japan, the Rokkasho Reprocessing Plant will be operated shortly after the year 2000, and the planning of the construction of the second commercial plant will be decided around 2010. Also, it is described that spent fuel storage has a positive meaning as an energy resource for the future utilization of Pu. Considering the balance between the increase of spent fuels and the domestic reprocessing capacity in Japan, it can be expected that the long-term storage of UO 2 spent fuels will be required. Then, we studied the effect of long-term storage of spent fuels on Pu-thermal fuel cycle. The burnup calculation were performed on the typical Japanese PWR fuel, and the burnup and criticality calculations were carried out on the Pu-thermal cores with MOX fuel. Based on the results, we evaluate the influence of extending the spent fuel storage term on the criticality safety, shielding design of the reprocessing plant and the core life time of the MOX core, etc. As the result of this work on long-term storage of LWR spent fuels, it becomes clear that there are few demerits regarding the lifetime of a MOX reactor core, and that there are many merits regarding the safety aspects of the fuel cycle facilities. Furthermore, long-term storage is meaningful as energy storage for effective utilization of Pu to be improved by technological innovation in future, and it will allow for sufficient time for the important policymaking of nuclear fuel cycle establishment in Japan. (author)

  18. Filtered atmospheric venting of LWR containments: the Swedish research programme and design concepts

    International Nuclear Information System (INIS)

    Graeslund, C.; Johansson, K.; Nilsson, L.; Tiren, I.

    1981-01-01

    An investigation of filtered atmospheric venting of LWR containments has been recommended by a governmental reactor safety committee as a means of reducing the large releases of radioactivity which it is believed could arise as a result of accidents beyond the design bases (class 9) in nuclear power plants. The purpose of the project is to provide the technical basis for evaluating the feasibility, effectiveness and costs of some vent filter design concepts. The main design objective is substantially to reduce releases to the atmosphere of those radioactive substances which could cause long-lasting contamination of large land areas resulting from accidents beyond the design basis. The degree to which that objective can be reached by applying vent-filter functions becomes the main design evaluation criterion. The governing principle for the vent filter design is to utilize passive components and functions to the greatest possible extent. The design concept is to vent the stream and gases from the containment into an underground tunnel containing a large bed of gravel where the steam is condensed. Non-condensable gases are vented through a sand filter at the outlet of the tunnel via a stack to the atmosphere. The tunnel volume envisaged is of the order of 100,000m 3 , and the length about 1000m. A deep tunnel in rock can be made to withstand the pressures from the burning of hydrogen-air mixtures. As an alternative method the condensing and filtering functions can be achieved by utilizing water pools built sub-surface in concrete structures. Concrete structures can also be built to withstand hydrogen burning. (author)

  19. Assessment of dry storage performance of spent LWR fuel assemblies with increasing burnup

    International Nuclear Information System (INIS)

    Peehs, M.; Garzarolli, F.; Goll, W.

    1999-01-01

    Although the safety of a dry long-term spent fuel store is scarcely influenced if a few fuel rods start to leak during extended storage - since all confinement systems are designed to retain gaseous activity safely - it is a very conservative safety goal to avoid the occurrence of systematic rod defects. To assess the extended storage performance of a spent fuel assembly (FA), the experience can be collated into 3 storage modes: I - fast rate of temperature decrease δ max ≥ δ ≥ 300 deg. C, II - medium rate of decrease for the fuel rod dry storage temperature 300 deg. C > δ ≥ 200 deg. C, III - slow to negligible rate of temperature decrease for δ 2 -fuel are practically immobile during storage. Consequently all fission-product-driven defect mechanisms will not take place. The leading defect mechanism - also for fuel rods with increased burnup - remains creep due to the hoop strain resulting from the fuel rod internal fission gas pressure. Limiting the creep to its primary and secondary stages prevents fuel rod degradation. The allowable uniform strain of the cladding is 1 - 2%. Calculations were performed to predict the dry storage performance of fuel assemblies with a burnup ≤ 55 GW · d/tHM based on the fuel assemblies end of life (EOL)-data and on a representative curve T = f(t). The maximum allowable hot spot temperature of a fuel rod in the CASTOR V cask was between 348 deg. C (U FA) and 358 deg. C (MOX FA). The highest hoop strain predicted after 40 years of storage is 0.77% proving that spent LWR fuel dry storage is safe. (author)

  20. Uncertainties in radioactivity release from LWR plants under LOCA conditions - magnitude and consequences

    International Nuclear Information System (INIS)

    Mattila, L.J.

    1977-01-01

    Standardized, deterministic, and supposedly conservative calculation methods and parameter values are applied in radiological safety analyses required for licensing individual nuclear power plants. As realistic as possible and comprehensive analyses are, however, absolutely necessary for many purposes, such as developing improved designs, comparisons between nuclear and non-nuclear power plant alternatives or entire energy production strategies, and also formulating improved acceptance criteria for plant licensing. A specific type of LOCA, called design basis accident (DBA), has obtained an exceptionally important status in the licensing procedure of light water reactor nuclear power plants. This postulated accident has a decisive influence on plant siting and on the design of the various engineered safety features. To avoid certain potential negative effects of the highly standardized guideline-based accident analysis procedure - such as introduction of apparent design ''improvements'', wrong priorization of research efforts, etc. - and to provide a realistic view about the safety of light water reactors to supplement the conservative results from regulatory analyses, a comprehensive understanding of the radiological consequences of LOCA's is indispensable. Estimates of fission product release from LWR plants under different LOCA conditions are associated with uncertainties due to deficient knowledge and truly random variability. The following steps of the fission product transport chain are discussed: generation of activity, fission product release from fuel to fuel pin voids prior to the accident, fuel rod puncturing and fission product release from punctured rods during the accident, further release from fuel during the transient, transport to the containment and finally removal in and leakage from the containment. Numerical examples are given by comparing assumptions, parameter values, and results from the following four analyses: the present guideline

  1. PLUTON, Isotope Generation and Depletion in Highly Irradiated LWR Fuel Rods

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Motoe, Suzuki

    2003-01-01

    1 - Description of program or function: The PLUTON-PC is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO 2 , UO 2 -Gd 2 O 3 , inhomogeneous MOX, and UO 2 -ThO 2 . The PLUTON-PC code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. 2 - Methods: Based upon cumulative yields, the PLUTON-PC code calculates as a function of radial position and local burnup concentrations of fission products, macroscopic scattering cross-sections and self-shielding effect which is important for standard fuel (for Pu-242 mainly) and more importantly for homogeneous and inhomogeneous MOX fuel because of higher concentrations of fissile and fertile isotopes of plutonium. The code results in burnup dependent fission rate density profiles throughout the in-reactor irradiation of LWR fuel rods. The isotopes included in calculations have been extended to cover all trans-uranium groups (plutonium plus higher actinides) of fissile and fertile isotopes. Self-shielding problem and scattering effects have been revised and solved for all isotopes in the calculations for adequacy at high burnup, different irradiation conditions and cladding materials

  2. Impacts of transient heat transfer modeling on prediction of advanced cladding fracture during LWR LBLOCA

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Youho, E-mail: euo@kaist.ac.kr; Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

    2016-03-15

    Highlights: • Use of constant heat transfer coefficient for fracture analysis is not sound. • On-time heat transfer coefficient should be used for thermal fracture prediction. • ∼90% of the actual fracture stresses were predicted with the on-time transient h. • Thermal-hydraulic codes can be used to better predict brittle cladding fracture. • Effects of surface oxides on thermal shock fracture should be accounted by h. - Abstract: This study presents the importance of coherency in modeling thermal-hydraulics and mechanical behavior of a solid for an advanced prediction of cladding thermal shock fracture. In water quenching, a solid experiences dynamic heat transfer rate evolutions with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates has been overlooked in the analysis of thermal shock fracture. In this study, we are presenting quantitative evidence against the prevailing use of a constant heat transfer coefficient for thermal shock fracture analysis in water. We conclude that no single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials. The presented results show a remarkable stress prediction improvement up to 80–90% of the actual stress with the use of the surface temperature dependent heat transfer coefficient. For thermal shock fracture analysis of brittle fuel cladding such as oxidized zirconium-based alloy or silicon carbide during LWR reflood, transient subchannel heat transfer coefficients obtained from a thermal-hydraulics code should be used as input for stress analysis. Such efforts will lead to a fundamental improvement in thermal shock fracture predictability over the current experimental empiricism for cladding fracture analysis during reflood.

  3. Program plan for research and development in support of LWR fuel recycle

    International Nuclear Information System (INIS)

    1975-01-01

    The ERDA program that is being planned to assist industry in the commercialization of the LWR fuel cycle will involve a range of activities, including joint programs with industry, R and D to provide technology, conceptual design of fuel recycle facilities, and environmental and economic assessments. A two-part program to begin in 1976 that is a portion of the overall ERDA plan is described. Responsibility for coordination and management of the tasks described in this document has been assigned to Du Pont as prime contractor to the ERDA Savannah River Operations Office. The first part of the program consists of the conceptual design of complete recycle facilities. The second part of the program, which will proceed concurrently, consists of supporting R and D activities, economic and environmental studies, and other studies to assist in the regulatory process. The R and D program will include both near-term activities in support of the conceptual design effort, and other activities aimed at general improvements in fuel cycle technology. The conceptual design will be used to develop current cost information for a complete reprocessing complex. The design will be based initially on current technology with provision for improvements as confirmatory information and advanced technology become available from the R and D program. The conceptual design and cost estimate will be developed by the Du Pont Atomic Energy Division. The R and D program and supporting studies will be directed at uncertainties in current technology as well as toward development of improved technology. It will include such R and D as might be appropriate for ERDA to undertake in support of joint programs with industry. The Savannah River Laboratory will have responsibility for coordinating the program

  4. Arguments for the installation of small LWR nuclear power plants in developing countries

    International Nuclear Information System (INIS)

    Lezenik, B.

    1977-01-01

    The differences disclosed in the IAEA-Report ''Market Survey...'' before and after the drastic increase of the oil price in 1973 indicate most unequivocally the growing economic profits of the application of nuclear power plants with smaller output. The report furthermore demonstrates in clear terms the benefits of a long-term fuel arrangement as is applied for nuclear power plants. Following these basic considerations, according to which the construction of nuclear power plants appears to be reasonable, economically speaking, the next important criterion having priority, as far as the technical aspect is concerned, is the question concerning the largest possible unit to be integrated into the grid. So as to ensure a steady mains power supply, the structure of operating power plant types, their number and unit sizes as well as their controllability and load-following behavior have to be coordinated with one another individually. As an example for the nuclear power plant Borssele (450 MW - PWR) which has been in operation since 1973, basic control concepts are outlined which together with the load-following behavior of this reactor meet the grid requirements. This is to demonstrate the flexible application of this reactor type. Manufacturing and operating experience as well as the accumulated know-how concerning design and construction of large plants in the 1200/1300 MW class offer a broad basis for the standardization of the smaller LWR-plants. The experience gained with proven components and approved systems for the fully standardized 1200/1300 MW reactors equally guarantees operating reliability and availability, since differences in output are realized by means of additive modifications of the primary system components. A comparison with, and an interpretation of, the essential design data outline the progress achieved in the standardization process. So as to ensure as efficient as possible implementation of the construction procedure, and the application of

  5. Bio-Based Coatings for Paper Applications

    Directory of Open Access Journals (Sweden)

    Vibhore Kumar Rastogi

    2015-11-01

    Full Text Available The barrier resistance and wettability of papers are commonly controlled by the application of petroleum-based derivatives such as polyethylene, waxes and/or fluor- derivatives as coating. While surface hydrophobicity is improved by employing these polymers, they have become disfavored due to limitations in fossil-oil resources, poor recyclability, and environmental concerns on generated waste with lack of biodegradation. Alternatively, biopolymers including polysaccharides, proteins, lipids and polyesters can be used to formulate new pathways for fully bio-based paper coatings. However, difficulties in processing of most biopolymers may arise due to hydrophilicity, crystallization behavior, brittleness or melt instabilities that hinder a full exploitation at industrial scale. Therefore, blending with other biopolymers, plasticizers and compatibilizers is advantageous to improve the coating performance. In this paper, an overview of barrier properties and processing of bio-based polymers and their composites as paper coating will be discussed. In particular, recent technical advances in nanotechnological routes for bio-based nano- composite coatings will be summarized, including the use of biopolymer nanoparticles, or nanofillers such as nanoclay and nanocellulose. The combination of biopolymers along with surface modification of nanofillers can be used to create hierarchical structures that enhance hydrophobicity, complete barrier protection and functionalities of coated papers.

  6. Agent Based Modeling Applications for Geosciences

    Science.gov (United States)

    Stein, J. S.

    2004-12-01

    Agent-based modeling techniques have successfully been applied to systems in which complex behaviors or outcomes arise from varied interactions between individuals in the system. Each individual interacts with its environment, as well as with other individuals, by following a set of relatively simple rules. Traditionally this "bottom-up" modeling approach has been applied to problems in the fields of economics and sociology, but more recently has been introduced to various disciplines in the geosciences. This technique can help explain the origin of complex processes from a relatively simple set of rules, incorporate large and detailed datasets when they exist, and simulate the effects of extreme events on system-wide behavior. Some of the challenges associated with this modeling method include: significant computational requirements in order to keep track of thousands to millions of agents, methods and strategies of model validation are lacking, as is a formal methodology for evaluating model uncertainty. Challenges specific to the geosciences, include how to define agents that control water, contaminant fluxes, climate forcing and other physical processes and how to link these "geo-agents" into larger agent-based simulations that include social systems such as demographics economics and regulations. Effective management of limited natural resources (such as water, hydrocarbons, or land) requires an understanding of what factors influence the demand for these resources on a regional and temporal scale. Agent-based models can be used to simulate this demand across a variety of sectors under a range of conditions and determine effective and robust management policies and monitoring strategies. The recent focus on the role of biological processes in the geosciences is another example of an area that could benefit from agent-based applications. A typical approach to modeling the effect of biological processes in geologic media has been to represent these processes in

  7. Au Based Nanocomposites Towards Plasmonic Applications

    Science.gov (United States)

    Panniello, A.; Curri, M. L.; Placido, T.; Reboud, V.; Kehagias, N.; Sotomayor Torres, C. M.; Mecerreyes, D.; Agostiano, A.; Striccoli, M.

    2010-06-01

    Incorporation of nano-sized metals in polymers can transfer their unique features to the host matrix, providing nanocomposite materials with improved optical, electric, magnetic and mechanical properties. In this work, colloidal Au nanorods have been incorporated into PMMA based random co-polymer, properly functionalized with amino groups and the optical and morphological properties of the resulting nanocomposite have been investigated by spectroscopic and AFM measurements. Au nanorods have demonstrated to preserve the plasmon absorption and to retain morphological features upon the incorporation, thus making the final metal modified polymer composite exploitable for the fabrication of plasmonic devices. The prepared nanocomposites have been then patterned by Nano Imprint Lithography technique in order to demonstrate the viability of the materials towards optical applications.

  8. Proceedings of the 2007 LWR Fuel Performance Meeting / TopFuel 2007 'Zero by 2010'

    International Nuclear Information System (INIS)

    2007-01-01

    ANS, ENS, AESJ and KNS are jointly organizing the 2007 International LWR Fuel Performance Meeting following the successful ENS TopFuel meeting held during 22-26 October, 2006 in Salamaca, Spain. Merging three premier nuclear fuel design and performance meetings: the ANS LWR Fuel Performance Meeting, the ENS TopFuel and Asian Water Reactor Fuel Performance Meeting (WRFPM) created this international meeting. The meeting will be held annually on a tri-annual rotational basis in USA, Asia, and Europe. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as performance experience in commercial and test reactors. The meeting excludes front end and back end fuel issues, however, it covers all front and/or back issues that impact fuel designs and performance

  9. Workshop on initiation of stress corrosion cracking under LWR conditions: Proceedings

    International Nuclear Information System (INIS)

    Nelson, J.L.; Cubicciotti, D.; Licina, G.J.

    1988-05-01

    A workshop titled ''Initiation of Stress Corrosion Cracking under LWR Conditions'' was held in Palo Alto, California on November 13, 1986, hosted by the Electric Power Research Institute. Participants were experts on the topic from nuclear steam supply and component manufacturers, public and private research laboratories, and university environments. Presentations included discussions on the definition of crack initiation, the effects of environmental and electrochemical variables on cracking susceptibility, and detection methods for the determination of crack initiation events and measurement of critical environmental and stress parameters. Examination of the questions related to crack initiation and its relative importance to the overall question of cracking of LWR materials from these perspectives provided inputs to EPRI project managers on the future direction of research efforts designed to prevent and control cracking. Thirteen reports have been cataloged separately

  10. Standard problem exercise to validate criticality codes for spent LWR fuel transport container calculations

    International Nuclear Information System (INIS)

    Whitesides, G.H.; Stephens, M.E.

    1984-01-01

    During the past two years, a Working Group established by the Organization for Economic Co-Operation and Development's Nuclear Energy Agency (OECD-NEA) has been developing a set of criticality benchmark problems which could be used to help establish the validity of criticality safety computer programs and their associated nuclear data for calculation of ksub(eff) for spent light water reactor (LWR) fuel transport containers. The basic goal of this effort was to identify a set of actual critical experiments which would contain the various material and geometric properties present in spent LWR transport contrainers. These data, when used by the various computational methods, are intended to demonstrate the ability of each method to accurately reproduce the experimentally measured ksub(eff) for the parameters under consideration

  11. Development and testing of standardized procedures and reference data for LWR surveillance

    International Nuclear Information System (INIS)

    McElroy, W.N.

    1979-02-01

    The resources and talents of many national and international organizations and laboratories, both governmental and industrial, are being used to establish analysis methods for predicting the embrittlement condition of light water reactor (LWR) primary systems. The exact interrelationships and responsibilities between those developing, understanding, combining, and applying state-of-the-art technology in dosimetry, metallurgy, and fracture mechanics for reactor systems analysis are being carefully reviewed and studied. This has resulted in a more comprehensive definition of the scope of new and updated ASTM standards required for the analysis and interpretation of LWR pressure vessel surveillance results. Fifteen new and updated ASTM standards have now been identified, together with a restructuring of the main interfaces between the individual standard practices, guides, and methods. The paper briefly discusses these standards and the initial results of multi-laboratory research work involved in their validation and calibration

  12. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

  13. EUV lithography for 30nm half pitch and beyond: exploring resolution, sensitivity, and LWR tradeoffs

    Science.gov (United States)

    Putna, E. Steve; Younkin, Todd R.; Chandhok, Manish; Frasure, Kent

    2009-03-01

    The International Technology Roadmap for Semiconductors (ITRS) denotes Extreme Ultraviolet (EUV) lithography as a leading technology option for realizing the 32nm half-pitch node and beyond. Readiness of EUV materials is currently one high risk area according to assessments made at the 2008 EUVL Symposium. The main development issue regarding EUV resist has been how to simultaneously achieve high sensitivity, high resolution, and low line width roughness (LWR). This paper describes the strategy and current status of EUV resist development at Intel Corporation. Data is presented utilizing Intel's Micro-Exposure Tool (MET) examining the feasibility of establishing a resist process that simultaneously exhibits <=30nm half-pitch (HP) L/S resolution at <=10mJ/cm2 with <=4nm LWR.

  14. Behaviour of LWR core materials under accident conditions. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-12-01

    At the invitation of the Government of the Russian Federation, following a proposal of the International Working Group on Water Reactor Fuel Performance and Technology, the IAEA convened a Technical Committee Meeting on Behaviour of LWR Core Materials Under Accident Conditions from 9 to 13 October 1995 in Dimitrovgrad to analyze and evaluate the behaviour of LWR core materials under accident conditions with special emphasis on severe accidents. In-vessel severe accidents phenomena were considered in detail, but specialized thermal hydraulic aspects as well as ex-vessel phenomena were outside the scope of the meeting. Forty participants representing eight countries attended the meeting. Twenty-three papers were presented and discussed during five sessions. Refs, figs, tabs

  15. Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere

    International Nuclear Information System (INIS)

    Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.

    1983-09-01

    Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur

  16. Investigation of LWR environmental effect on fatigue lifetime of austenitic stainless steel component

    International Nuclear Information System (INIS)

    Kim, J. S.; Youm, H. K.; Jin, T. E.

    1999-01-01

    The fatigue lifetime of principal components in nuclear power plant is evaluated by using the design fatigue curves in ASME B and PV code during design process. However, it is inadequate to evaluate fatigue lifetime considering the LWR environmental effect by these design fatigue curves because these are presented only under atmosphere environment. Therefore, many studies are recently performed for the design fatigue curves considering LWR environmental effect and are presented that the design fatigue curves in ASME B and PV code can be non-conservative. In present paper, the limits and differences of the design fatigue curves considering environmental effect are presented. To investigate the change of fatigue lifetime according to each design fatigue curve, the CUFs for the pressurizer spray nozzle partly composed of austenitic stainless steel are calculated according to each one. Finally, if the evaluation result can not be satisfied with fatigue design requirement, the alternatives to reduce design cumulative usage factor are discussed. (author)

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

    International Nuclear Information System (INIS)

    Yamada, Ichita; Suzuki, Shigemitsu

    2008-01-01

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

  18. EUR, an European utility requirements documents for future LWR power stations

    International Nuclear Information System (INIS)

    Berbey, Pierre; Lienard, Michel; Redon, Ramon; Essmann, Juergen; Taylor, David T.

    2004-01-01

    A group of the major European utilities are developing a common requirement document which will be used for the LWR nuclear power plants to be built in Europe from the beginning of the next century. This document provides harmonised policies and technical requirements that will allow the implementation of a design developed in one country into another one. The objectives and contents of the document, the organisation set up for its production and the main requirements are summarised in the paper. (author)

  19. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    International Nuclear Information System (INIS)

    Sample, C.R.

    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

  20. 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.

  1. Comment: collection of assay data on isotopic composition in LWR spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Naito, Yoshitaka; Kurosawa, Masayoshi; Suyama, Kenya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Many assay data of LWR spent fuels have been collected from reactors in the world and some of them are already stored in the database SFCOMPO which was constructed on a personal computer IBM PC/AT. On the other hand, Group constant libraries for burnup calculation code ORIGEN-II were generated from the nuclear data file JENDL3.2. These libraries were evaluated by using the assay data in SFCOMPO. (author)

  2. Minutes of the Twelfth LWR pressure vessel surveillance dosimtery improvement program meeting

    International Nuclear Information System (INIS)

    1989-01-01

    The 1983 Twelfth Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP) Meeting, which was held October 24-28, 1983. Sections 1 through 14 of this report provide documentation of agreements, commitments, and reports that are subject to the approval and concurrence of the participating laboratories and supporting agencies and organizations. Attachment No. 1 provides information on the preparation of a number of NUREG publications that will document the results of various aspects of the LWR-PV-SDIP. For each NUREG publication, a tentative ''Table of Contents'' is provided in addition to suggested interlaboratory writing assignments and camera-ready copy contribution due dates, as appropriate. Attachment No. 2 provides information on planning for the Fifth ASTM-EURATOM Symposium. Attachment No. 3 provides information on an ASTM press release about an MPC-6 meeting and dpa and E > 1 MeV exposure parameters. Attachments No. 4 and 5 provide copies of two LWR-PV-SDIP related papers presented at the Eleventh WRSR Information Meeting, October 24-28, 1983

  3. APEX nuclear fuel cycle for production of LWR fuel and elimination of radioactive waste

    International Nuclear Information System (INIS)

    Steinberg, M.; Powell, J.R.

    1981-08-01

    The development of a nuclear fission fuel cycle is proposed which eliminates all the radioactive fission product waste effluent and the need for geological-age high level waste storage and provides a long term supply of fissile fuel for an LWR power reactor economy. The fuel cycle consists of reprocessing LWR spent fuel (1 to 2 years old) to remove the stable nonradioactive (NRFP, e.g. lanthanides, etc.) and short-lived fission products SLFP e.g. half-lives of (1 to 2 years) and returning, in dilute form, the long-lived fission products, ((LLFPs, e.g. 30 y half-life Cs, Sr, and 10 y Kr, and 16 x 10 6 y I) and the transuranics (TUs, e.g. Pu, Am, Cm, and Np) to be refabricated into fresh fuel elements. Makeup fertile and fissile fuel are to be supplied through the use of a Spallator (linear accelerator spallation-target fuel-producer). The reprocessing of LWR fuel elements is to be performed by means of the Chelox process which consists of Airox treatment (air oxidation and hydrogen reduction) followed by chelation with an organic reagent (β-diketonate) and vapor distillation of the organometallic compounds for separation and partitioning of the fission products

  4. Review of provisions on corrosion fatigue and stress corrosion in WWER and Western LWR Codes and Standards

    International Nuclear Information System (INIS)

    Buckthorpe, D.; Filatov, V.; Tashkinov, A.; Evropin, S.V.; Matocha, K.; Guinovart, J.

    2003-01-01

    Results are presented from a collaborative project performed on behalf of the European Commission, Working Group Codes and Standards. The work covered the contents of current codes and standards, plant experience and R and D results. Current fatigue design rules use S-N curves based on tests in air. Although WWER and LWR design curves are often similar they are derived, presented and used in different ways and it is neither convenient nor appropriate to harmonise them. Similarly the fatigue crack growth laws used in the various design and in-service inspection rules differ significantly with respect to both growth rates in air and the effects of water reactor environments. Harmonised approaches to the effects of WWER and LWR environments are possible based on results from R and D programmes carried out over the last decade. For carbon and low alloy steels a consistent approach to both crack initiation and growth can be formulated based on the superposition of environmentally assisted cracking effects on the fatigue crack development. The approach indicates that effects of the water environment are minimal given appropriate control of the oxygen content of the water and/or the sulphur content of the steel. For austenitic stainless steels a different mechanisms may apply and a harmonised approach is possible at present only for S-N curves. Although substantial progress has been made with respect to corrosion fatigue, more data and a clearer understanding are required in order to write code provisions particularly in the area of high cycle fatigue. Reactor operation experience shows stress corrosion cracking of austenitic steels is the most common cause of failure. These failures are associated with high residual stresses combined with high levels of dissolved oxygen or the presence of contaminants. For primary circuit internals there is a potential threat to integrity from irradiated assisted stress corrosion cracking. Design and in-service inspection rules do not at

  5. Measurement of blockage in deformed LWR multi-rod arrays

    International Nuclear Information System (INIS)

    Hindle, E.D.; Jones, C.; Whitty, S.

    1983-01-01

    This paper critically reviews the current methods used for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed. Also examples of the application of automatic computerised techniques to directly measure rod strain, blockage, sub-channel blockage and perimeter changes from photographs of sections through deformed arrays are presented. (author)

  6. Proceeding of 35th domestic symposium on applications of structural reliability and risk assessment methods to nuclear power plants

    International Nuclear Information System (INIS)

    2005-06-01

    As the 35th domestic symposium of Atomic Energy Research Committee, the Japan Welding Engineering Society, the symposium was held titled as Applications of structural reliability/risk assessment methods to nuclear energy'. Six speakers gave lectures titled as 'Structural reliability and risk assessment methods', 'Risk-informed regulation of US nuclear energy and role of probabilistic risk assessment', 'Reliability and risk assessment methods in chemical plants', 'Practical structural design methods based on reliability in architectural and civil areas', 'Maintenance activities based on reliability in thermal power plants' and 'LWR maintenance strategies based on Probabilistic Fracture Mechanics'. (T. Tanaka)

  7. Yeast-based biosensors: design and applications.

    Science.gov (United States)

    Adeniran, Adebola; Sherer, Michael; Tyo, Keith E J

    2015-02-01

    Yeast-based biosensing (YBB) is an exciting research area, as many studies have demonstrated the use of yeasts to accurately detect specific molecules. Biosensors incorporating various yeasts have been reported to detect an incredibly large range of molecules including but not limited to odorants, metals, intracellular metabolites, carcinogens, lactate, alcohols, and sugars. We review the detection strategies available for different types of analytes, as well as the wide range of output methods that have been incorporated with yeast biosensors. We group biosensors into two categories: those that are dependent upon transcription of a gene to report the detection of a desired molecule and those that are independent of this reporting mechanism. Transcription-dependent biosensors frequently depend on heterologous expression of sensing elements from non-yeast organisms, a strategy that has greatly expanded the range of molecules available for detection by YBBs. Transcription-independent biosensors circumvent the problem of sensing difficult-to-detect analytes by instead relying on yeast metabolism to generate easily detected molecules when the analyte is present. The use of yeast as the sensing element in biosensors has proven to be successful and continues to hold great promise for a variety of applications. © FEMS 2015. All rights reserved. For permissions, please e-mail: journals.permission@oup.com.

  8. Investigation of nuclide importance to functional requirements related to transport and long-term storage of LWR spent fuel

    International Nuclear Information System (INIS)

    Broadhead, B.L.; DeHart, M.D.; Ryman, J.C.; Tang, J.S.; Parks, C.V.

    1995-06-01

    This study investigates the relative importances of the various actinide, fission-product, and light-element isotopes associated with LWR spent fuel with respect to five analysis areas: criticality safety (absorption fractions), shielding (dose rate fractions), curies (fractional curies levels), decay heat (fraction of total watts), and radiological toxicity (fraction of potential committed effective dose equivalent). These rankings are presented for up to six different burnup/enrichment scenarios and at decay times from 2 to 100,000 years. Ranking plots for each of these analysis areas are given in an Appendix for completeness, as well as summary tables in the main body of the report. Summary rankings are presented in terms of high (greater than 10% contribution to the total), medium (between 1% and 10% contribution), and low (less than 1% contribution) for both short- and long-term cooling. When compared with the expected measurement accuracies, these rankings show that most of the important isotopes can be characterized sufficiently for the purpose of radionuclide generation/depletion code validation in each of the analysis areas. Because the main focus of this work is on the relative importances of isotopes associated with L at sign spent fuel, some conclusions may not be applicable to similar areas such as high-level waste (HLW) and nonfuel-bearing components (NFBC)

  9. Development of LWR Fuels with Enhanced Accident Tolerance

    International Nuclear Information System (INIS)

    Lahoda, Edward J.; Boylan, Frank A.

    2015-01-01

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company's Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15 N and U 3 Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3 Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3 Si 2 will represent a 15% increase in U235 loadings over those in UO fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3 Si 2 /68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2 , UN, and U 3 Si 2 the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2 . Cold spray application of either the amorphous steel or the Ti 2 AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the

  10. Development of LWR Fuels with Enhanced Accident Tolerance

    Energy Technology Data Exchange (ETDEWEB)

    Lahoda, Edward J. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States); Boylan, Frank A. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States)

    2015-10-30

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing

  11. ANDROID CHATTING APPLICATION BASED ON WI-FI TECHNOLOGY

    OpenAIRE

    Ekata M. Lambture *, Prof. Z.M. Shaikh

    2016-01-01

    As we know that Android operating system is one of the popular operating system for mobile applications.It supports mobile applications such as Chatting based applications,File sharing applications etc. In market thosands of applications are available in which users can send-receive messages,shares the files,send images,audio,video through network connectivity such as 2G,3G,4G. Android chatting application based on wi-fi technology provides same features but using WI-FI(Wireless Fidelity)...

  12. Service quality of cloud-based applications

    CERN Document Server

    Bauer, Eric

    2014-01-01

    This book explains why applications running on cloud might not deliver the same service reliability, availability, latency and overall quality to end users as they do when the applications are running on traditional (non-virtualized, non-cloud) configurations, and explains what can be done to mitigate that risk.

  13. Advanced applications of scatterometry based optical metrology

    Science.gov (United States)

    Dixit, Dhairya; Keller, Nick; Kagalwala, Taher; Recchia, Fiona; Lifshitz, Yevgeny; Elia, Alexander; Todi, Vinit; Fronheiser, Jody; Vaid, Alok

    2017-03-01

    The semiconductor industry continues to drive patterning solutions that enable devices with higher memory storage capacity, faster computing performance, and lower cost per transistor. These developments in the field of semiconductor manufacturing along with the overall minimization of the size of transistors require continuous development of metrology tools used for characterization of these complex 3D device architectures. Optical scatterometry or optical critical dimension (OCD) is one of the most prevalent inline metrology techniques in semiconductor manufacturing because it is a quick, precise and non-destructive metrology technique. However, at present OCD is predominantly used to measure the feature dimensions such as line-width, height, side-wall angle, etc. of the patterned nano structures. Use of optical scatterometry for characterizing defects such as pitch-walking, overlay, line edge roughness, etc. is fairly limited. Inspection of process induced abnormalities is a fundamental part of process yield improvement. It provides process engineers with important information about process errors, and consequently helps optimize materials and process parameters. Scatterometry is an averaging technique and extending it to measure the position of local process induced defectivity and feature-to-feature variation is extremely challenging. This report is an overview of applications and benefits of using optical scatterometry for characterizing defects such as pitch-walking, overlay and fin bending for advanced technology nodes beyond 7nm. Currently, the optical scatterometry is based on conventional spectroscopic ellipsometry and spectroscopic reflectometry measurements, but generalized ellipsometry or Mueller matrix spectroscopic ellipsometry data provides important, additional information about complex structures that exhibit anisotropy and depolarization effects. In addition the symmetry-antisymmetry properties associated with Mueller matrix (MM) elements

  14. Arduino based acquisition system for control applications

    OpenAIRE

    Granado Navarro, Miguel Ángel

    2012-01-01

    The aim of this project is to design an implement a low cost acquisition system intended for control applications using the Arduino prototyping platform. Arduino has become a popular open-source single-board microcontroller among electronic hobbyists, and it is gaining acceptance as a quick prototyping tool for engineering and educational projects also. A computer, using a control application done in LabVIEW or IDE Arduino, must control all the system by cable or wireless. The pro...

  15. Electrospinning cellulose based nanofibers for sensor applications

    Science.gov (United States)

    Nartker, Steven

    2009-12-01

    Bacterial pathogens have recently become a serious threat to the food and water supply. A biosensor based on an electrochemical immunoassay has been developed for detecting food borne pathogens, such as Escherichia coli (E. coli) O157:H7. These sensors consist of several materials including, cellulose, cellulose nitrate, polyaniline and glass fibers. The current sensors have not been optimized in terms of microscale architecture and materials. The major problem associated with the current sensors is the limited concentration range of pathogens that provides a linear response on the concentration conductivity chart. Electrospinning is a process that can be used to create a patterned fiber mat design that will increase the linear range and lower the detection limit of these sensors by improving the microscale architecture. Using the electrospinning process to produce novel mats of cellulose nitrate will offer improved surface area, and the cellulose nitrate can be treated to further improve chemical interactions required for sensor activity. The macro and micro architecture of the sensor is critical to the performance of the sensors. Electrospinning technology can be used to create patterned architectures of nanofibers that will enhance sensor performance. To date electrospinning of cellulose nitrate has not been performed and optimization of the electrospinning process will provide novel materials suitable for applications such as filtration and sensing. The goal of this research is to identify and elucidate the primary materials and process factors necessary to produce cellulose nitrate nanofibers using the electrospinning process that will improve the performance of biosensors. Cellulose nitrate is readily dissolved in common organic solvents such as acetone, tetrahydrofuran (THF) and N,N dimethylformamide (DMF). These solvents can be mixed with other latent solvents such as ethanol and other alcohols to provide a solvent system with good electrospinning behavior

  16. Value-based medicine: concepts and application

    OpenAIRE

    Jong-Myon Bae

    2015-01-01

    Global healthcare in the 21st century is characterized by evidence-based medicine (EBM), patient-centered care, and cost effectiveness. EBM involves clinical decisions being made by integrating patient preference with medical treatment evidence and physician experiences. The Center for Value-Based Medicine suggested value-based medicine (VBM) as the practice of medicine based upon the patient-perceived value conferred by an intervention. VBM starts with the best evidence-based data and conver...

  17. Storage of LWR spent fuel in air: Volume 1: Design and operation of a spent fuel oxidation test facility

    International Nuclear Information System (INIS)

    Thornhill, C.K.; Campbell, T.K.; Thornhill, R.E.

    1988-12-01

    This report describes the design and operation and technical accomplishments of a spent-fuel oxidation test facility at the Pacific Northwest Laboratory. The objective of the experiments conducted in this facility was to develop a data base for determining spent-fuel dry storage temperature limits by characterizing the oxidation behavior of light-water reactor (LWR) spent fuels in air. These data are needed to support licensing of dry storage in air as an alternative to spent-fuel storage in water pools. They are to be used to develop and validate predictive models of spent-fuel behavior during dry air storage in an Independent Spent Fuel Storage Installation (ISFSI). The present licensed alternative to pool storage of spent fuel is dry storage in an inert gas environment, which is called inerted dry storage (IDS). Licensed air storage, however, would not require monitoring for maintenance of an inert-gas environment (which IDS requires) but does require the development of allowable temperature limits below which UO 2 oxidation in breached fuel rods would not become a problem. Scoping tests at PNL with nonirradiated UO 2 pellets and spent-fuel fragment specimens identified the need for a statistically designed test matrix with test temperatures bounding anticipated maximum acceptable air-storage temperatures. This facility was designed and operated to satisfy that need. 7 refs

  18. Analysis of reactivity worths of highly-burnt PWR fuel samples measured in LWR-PROTEUS Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Grimm, Peter; Murphy, Michael F.; Jatuff, Fabian; Seiler, Rudolf [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland)

    2008-07-01

    The reactivity loss of PWR fuel with burnup has been determined experimentally by inserting fresh and highly-burnt fuel samples in a PWR test lattice in the framework of the LWR-PROTEUS Phase II programme. Seven UO{sub 2} samples irradiated in a Swiss PWR plant with burnups ranging from approx40 to approx120 MWd/kg and four MOX samples with burnups up to approx70 MWd/kg were oscillated in a test region constituted of actual PWR UO{sub 2} fuel rods in the centre of the PROTEUS zero-power experimental facility. The measurements were analyzed using the CASMO-4E fuel assembly code and a cross section library based on the ENDF/B-VI evaluation. The results show close proximity between calculated and measured reactivity effects and no trend for a deterioration of the quality of the prediction at high burnup. The analysis thus demonstrates the high accuracy of the calculation of the reactivity of highly-burnt fuel. (authors)

  19. CRISSUE-S, Neutronics/Thermal-hydraulics Coupling in LWR Technology

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Bousbia Salah, Anis; Galassi, G.M.; Vedovi, Juswald; Van Goethem, Georges; Hadek, Jan; Macek, Jiri; Rindelhardt, Udo; Rohde, Ulrich; Ahnert Iglesias, Carol; Aragones Beltran, Jose Maria; Reventos, Francesc; Cuadra, Arantxa; Gago, Jose Luis; Verdu, Gumersindo; Miro, Rafael; Ginestar, Damian; Sanchez, Ana Maria; Sjoberg, Anders; Yitbarek, M.; Sandervag, Oddbjoern; Garis, Ninos; Frid, Wiktor; Panayotov, Dobromir; Ivanov, Kostadin; Uddin, Rizwan; Sartori, Enrico

    2004-01-01

    expertise and findings and gathering together expert scientists from various areas relevant to the issues addressed. Added value for the CRISSUE-S activity consists of proposing and making available a list of transients to be analysed by coupled neutron kinetics/thermal-hydraulic techniques and of defining ?acceptability? (or required precision) thresholds for the results of the analyses. The list of transients is specific to the different NPP types such as PWR, BWR and VVER. The acceptability thresholds for calculation precision are general in nature and are applicable to all LWRs. The creation of a database including the main results from coupled 3-D neutron kinetics/thermal-hydraulic calculations and their analysis should also be noted. The CRISSUE-S project is organised into three work packages (WPs). The first WP includes activities related to obtaining and documenting relevant data. The second WP is responsible for the state-of-the-art report (SOAR), while the third WP concerns the evaluation of the findings from the SOAR and includes outcomes of the entire project formulated as recommendations, mainly to the nuclear power industry and to the regulatory authorities. The present report is the result of the first WP and discusses the type of transients that are of interest in relation to reactivity-initiated accidents in LWR NPPs and elaborates on the data required for coupled 3-D neutron kinetics/thermal hydraulic analysis and also on data needed to perform associated validations. The reports have been written to accomplish the objectives established in the contract between the EU and its partners. Expected beneficiaries include institutions and organisations involved with nuclear technology (e.g. utilities, regulators, research, fuel industry). In addition, specific expected beneficiaries are junior- or senior-level researchers and technologists working in the considered field of research and development and application of coupled neutron kinetics/thermal-hydraulics

  20. Total sensitivity and uncertainty analysis for LWR pin-cells with improved UNICORN code

    International Nuclear Information System (INIS)

    Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Shen, Wei

    2017-01-01

    Highlights: • A new model is established for the total sensitivity and uncertainty analysis. • The NR approximation applied in S&U analysis can be avoided by the new model. • Sensitivity and uncertainty analysis is performed to PWR pin-cells by the new model. • The effects of the NR approximation for the PWR pin-cells are quantified. - Abstract: In this paper, improvements to the multigroup cross-section perturbation model have been proposed and applied in the self-developed UNICORN code, which is capable of performing the total sensitivity and total uncertainty analysis for the neutron-physics calculations by applying the direct numerical perturbation method and the statistical sampling method respectively. The narrow resonance (NR) approximation was applied in the multigroup cross-section perturbation model, implemented in UNICORN. As improvements to the NR approximation to refine the multigroup cross-section perturbation model, an ultrafine-group cross-section perturbation model has been established, in which the actual perturbations are applied to the ultrafine-group cross-section library and the reconstructions of the resonance cross sections are performed by solving the neutron slowing-down equation. The total sensitivity and total uncertainty analysis were then applied to the LWR pin-cells, using both the multigroup and the ultrafine-group cross-section perturbation models. The numerical results show that the NR approximation overestimates the relative sensitivity coefficients and the corresponding uncertainty results for the LWR pin-cells, and the effects of the NR approximation are significant for σ_(_n_,_γ_) and σ_(_n_,_e_l_a_s_) of "2"3"8U. Therefore, the effects of the NR approximation applied in the total sensitivity and total uncertainty analysis for the neutron-physics calculations of LWR should be taken into account.

  1. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

    Directory of Open Access Journals (Sweden)

    Saas Laurent

    2017-01-01

    Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

  2. Projection of US LWR spent fuel storage requirements

    International Nuclear Information System (INIS)

    Fletcher, J.F.; Cole, B.M.; Purcell, W.L.; Rau, R.G.

    1982-11-01

    The spent fuel storage requirements projection is based on data supplied for each operating or planned nuclear power power plant by the operting utilities. The data supplied by the utilities encompassed details of plant operating history, past records of fuel discharges, current inventories in reactor spent fuel storage pools, and projections of future discharge patterns. Data on storage capacity of storage pools and on characterization of the discharged fuel are also included. The data supplied by the utilities, plus additional data from other appropriate sources, are maintained on a computerized data base by Pacific Northwest Laboratory. The spent fuel requirements projection was based on utility data updated and verified as of December 31, 1981

  3. PANDA: A Multipurpose Integral Test Facility for LWR Safety Investigations

    International Nuclear Information System (INIS)

    Paladino, D.; Dreier, J.

    2012-01-01

    The PANDA facility is a large scale, multicompartmental thermal hydraulic facility suited for investigations related to the safety of current and advanced LWRs. The facility is multipurpose, and the applications cover integral containment response tests, component tests, primary system tests, and separate effect tests. Experimental investigations carried on in the PANDA facility have been embedded in international projects, most of which under the auspices of the EU and OECD and with the support of a large number of organizations (regulatory bodies, technical dupport organizations, national laboratories, electric utilities, industries) worldwide. The paper provides an overview of the research programs performed in the PANDA facility in relation to BWR containment systems and those planned for PWR containment systems.

  4. Aging of reactor vessels in LWR type reactors

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-01-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs

  5. Evaluating the loss of a LWR spent fuel or plutonium shipping package into the sea

    International Nuclear Information System (INIS)

    Heaberlin, S.W.; Baker, D.A.

    1976-06-01

    As the nations of the world turn to nuclear power for an energy source, commerce in nuclear fuel cycle materials will increase. Some of this commerce will be transported by sea. Such shipments give rise to the possibility of loss of these materials into the sea. This paper discusses the postulated accidental loss of two materials, light water reactor (LWR) spent fuel and plutonium, at sea. The losses considered are that of a single shipping package which is either undamaged or damaged by fire prior to the loss. The containment failure of the package in the sea,

  6. Safety-related investigations on power distribution in MOX fuel elements in LWR cores

    International Nuclear Information System (INIS)

    Kramer, E.; Langenbuch, S.

    1991-01-01

    For the concept of thermal recycling various fuel assembly designs have been developped during the last years. An overview is given describing the present status of MOX-fuel assembly design for PWR and BWR. The local power distribution within the MOX-fuel assembly and influences between neighbouring MOX- and Uranium fuel assemblies have been analyzed by own calculations. These investigations are limited to specific aspects of the spatial power distribution, which are related to the use of MOX-fuel assemblies within the reactor core of LWR. (orig.) [de

  7. Residual life assessment of major LWR components: NPAR approach and results

    International Nuclear Information System (INIS)

    Shah, V.N.; Weidenhamer, G.H.; Vora, J.P.

    1991-01-01

    The nuclear plant aging research (NPAR) program is systematically addressing the technical issues associated with understanding and managing aging of major LWR components. Twenty-one major components have been identified and prioritized according to their relevance to plant safety. Qualitative aging assessment has identified pertinent design features, materials, stressors, environments, aging mechanisms. and failure modes for each of the components. Emerging inspection, surveillance, and monitoring methods to characterize aging damage and mitigation methods to reduce the damage are currently being assessed. The results of all these assessments are used to develop life-assessment procedures for the components and are included in appropriate documents supporting the regulatory requirements for license renewal. (author)

  8. Prioritization of tasks in the draft LWR safety technology program plan. Final report

    International Nuclear Information System (INIS)

    Lim, E.Y.; Miller, W.J.; Parkinson, W.J.; Ritzman, R.L.; vonHerrmann, J.L.; Wood, P.J.

    1980-05-01

    The purpose of this report is to describe both the approach taken and the results produced in the SAI effort to prioritize the tasks in the Sandia draft LWR Safety Technology Program Plan. This work used the description of important safety issues developed in the Reactor Safety Study (2) to quantify the effect of safety improvements resulting from a research and development program on the risk from nuclear power plants. Costs of implementation of these safety improvements were also estimated to allow a presentation of the final results in a value (i.e., risk reduction) vs. impact (i.e., implementation costs) matrix

  9. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    International Nuclear Information System (INIS)

    Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

    1984-01-01

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

  10. Advantages of retrofitting high velocity separators to LWR turbines; experience in VVR NPP Loviisa

    International Nuclear Information System (INIS)

    Dueymes, E.; Peyrelongue, J.P.

    1992-01-01

    Erosion-corrosion by wet steam is a concern for VVER operators and also, in numerous LWR power plants of western technology. The backfitting of moisture separators at the HP Turbine outlets is a way to avoid maintenance costs, repairs, replacement of pipes or equipments. Installation of HVS at LOVIISA confirms that this device, whose installation work is reduced to a minimum, is able to remove quite all the water from the steam just a few meters downstream the HP cylinder. A long term operation can be expected for carbon steel equipments, even those previously damaged by erosion-corrosion. (authors). 6 figs., 2 tabs

  11. Uncertainties in criticality analysis which affect the storage and transportation of LWR fuel

    International Nuclear Information System (INIS)

    Napolitani, D.G.

    1989-01-01

    Satisfying the design criteria for subcriticality with uncertainties affects: the capacity of LWR storage arrays, maximum allowable enrichment, minimum allowable burnup and economics of various storage options. There are uncertainties due to: calculational method, data libraries, geometric limitations, modelling bias, the number and quality of benchmarks performed and mechanical uncertainties in the array. Yankee Atomic Electric Co. (YAEC) has developed and benchmarked methods to handle: high density storage rack designs, pin consolidation, low density moderation and burnup credit. The uncertainties associated with such criticality analysis are quantified on the basis of clean criticals, power reactor criticals and intercomparison of independent analysis methods

  12. Synthesis and Application of Graphene Based Nanomaterials

    Science.gov (United States)

    Peng, Zhiwei

    Graphene, a two-dimensional sp2-bonded carbon material, has recently attracted major attention due to its excellent electrical, optical and mechanical properties. Depending on different applications, graphene and its derived hybrid nanomaterials can be synthesized by either bottom-up chemical vapor deposition (CVD) methods for electronics, or various top-down chemical reaction methods for energy generation and storage devices. My thesis begins with the investigation of CVD synthesis of graphene thin films in Chapter 1, including the direct growth of bilayer graphene on insulating substrates and synthesis of "rebar graphene": a hybrid structure with graphene and carbon or boron nitride nanotubes. Chapter 2 discusses the synthesis of nanoribbon-shaped materials and their applications, including splitting of vertically aligned multi-walled carbon nanotube carpets for supercapacitors, synthesis of dispersable ferromagnetic graphene nanoribbon stacks with enhanced electrical percolation properties in magnetic field, graphene nanoribbon/SnO 2 nanocomposite for lithium ion batteries, and enhanced electrocatalysis for hydrogen evolution reactions from WS2 nanoribbons. Next, Chapter 3 discusses graphene coated iron oxide nanomaterials and their use in energy storage applications. Finally, Chapter 4 introduces the development, characterization, and fabrication of laser induced graphene and its application as supercapacitors.

  13. Knowledge-Based Approaches: Two cases of applicability

    DEFF Research Database (Denmark)

    Andersen, Tom

    1997-01-01

    Basic issues of the term: A knowledge-based approach (KBA) are discussed. Two cases of applicable to KBA are presented, and its concluded that KBA is more than just IT.......Basic issues of the term: A knowledge-based approach (KBA) are discussed. Two cases of applicable to KBA are presented, and its concluded that KBA is more than just IT....

  14. Web-based control application using WebSocket

    International Nuclear Information System (INIS)

    Furukawa, Y.

    2012-01-01

    The WebSocket allows asynchronous full-duplex communication between a Web-based (i.e. Java Script-based) application and a Web-server. WebSocket started as a part of HTML5 standardization but has now been separated from HTML5 and has been developed independently. Using WebSocket, it becomes easy to develop platform independent presentation layer applications for accelerator and beamline control software. In addition, a Web browser is the only application program that needs to be installed on client computer. The WebSocket-based applications communicate with the WebSocket server using simple text-based messages, so WebSocket is applicable message-based control system like MADOCA, which was developed for the SPring-8 control system. A simple WebSocket server for the MADOCA control system and a simple motor control application were successfully made as a first trial of the WebSocket control application. Using Google-Chrome (version 13.0) on Debian/Linux and Windows 7, Opera (version 11.0) on Debian/Linux and Safari (version 5.0.3) on Mac OS X as clients, the motors can be controlled using a WebSocket-based Web-application. Diffractometer control application use in synchrotron radiation diffraction experiment was also developed. (author)

  15. Short Communication on "In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures"

    Science.gov (United States)

    Miao, Yinbin; Harp, Jason; Mo, Kun; Bhattacharya, Sumit; Baldo, Peter; Yacout, Abdellatif M.

    2017-02-01

    The radiation-induced amorphization of U3Si2 was investigated by in-situ transmission electron microscopy using 1 MeV Kr ion irradiation. Both arc-melted and sintered U3Si2 specimens were irradiated at room temperature to confirm the similarity in their responses to radiation. The sintered specimens were then irradiated at 350 °C and 550 °C up to 7.2 × 1015 ions/cm2 to examine their amorphization behavior under light water reactor (LWR) conditions. U3Si2 remains crystalline under irradiation at LWR temperatures. Oxidation of the material was observed at high irradiation doses.

  16. Technical program to study the benefits of nonlinear analysis methods in LWR component designs. Technical report TR-3723-1

    International Nuclear Information System (INIS)

    Raju, P.P.

    1980-05-01

    This report summarizes the results of the study program to assess the benefits of nonlinear analysis methods in Light Water Reactor (LWR) component designs. The current study reveals that despite its increased cost and other complexities, nonlinear analysis is a practical and valuable tool for the design of LWR components, especially under ASME Level D service conditions (faulted conditions) and it will greatly assist in the evaluation of ductile fracture potential of pressure boundary components. Since the nonlinear behavior is generally a local phenomenon, the design of complex components can be accomplished through substructuring isolated localized regions and evaluating them in detail using nonlinear analysis methods

  17. A Framework for Blockchain-Based Applications

    OpenAIRE

    Feig, Ephraim

    2018-01-01

    Blockchains have recently generated explosive interest from both academia and industry, with many proposed applications. But descriptions of many these proposals are more visionary projections than realizable proposals, and even basic definitions are often missing. We define "blockchain" and "blockchain network", and then discuss two very different, well known classes of blockchain networks: cryptocurrencies and Git repositories. We identify common primitive elements of both and use them to c...

  18. Status of LWR primary pressure boundary structural materials

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Byun, Taek Sang; Kang, Sung Sik; Ryu, Woo Seog; Lee, Bong Sang; Kook, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    The integrity of major systems, structures and components is a prerequisite to the economy and safety of an existing light water reactor and also for the next generation reactors. As few reactor structural materials are being manufactured by domestic companies, based on economic and safety reasons, a new demand to improve the quality of domestic reactor structural materials and to develop reactor structural steels has arisen. Investigations on the state-of-the-art of the materials specifications, performance and current state of structural materials development were performed as a first step to domestic reactor structural steel development and summarized the result in the present report. (Author) 10 refs., 10 figs., 21 tabs.

  19. Mechanistic prediction of iodine and cesium release from LWR fuel

    International Nuclear Information System (INIS)

    Rest, J.

    1983-12-01

    A theoretical model (FASTGRASS) has been used for predicting the behavior of fission gas and volatile fission products (VFPs) in UO 2 -base fuels during steady-state and transient conditions. This model represents an attempt to develop an efficient predictive capability for the full range of possible reactor operating conditions. Fission products released from the fuel are assumed to reach the fuel surface by successively diffusing (via atomic and gas-bubble mobility) from the grains to grain faces and then to the grain edges, where the fission products are released through a network of interconnected tunnels of fission-gas-induced and fabricated porosity

  20. Method for detecting resin leakage in LWR coolant

    International Nuclear Information System (INIS)

    Girard, J.E.

    1988-05-01

    Resin leakage from condensate polishing units can result in steam generator corrosion. This report describes the development of a resin leakage detection method based in analyzing the organic breakdown products released from resin on heating. The breakdown products are analyzed using high performance liquid chromatography (HPLC) with fluorescence detection. Some of the organic products formed have been identified. A design for a resin monitoring unit, suitable for incorporation into the IONTRAC system, is presented. Theoretically, detection of ppB levels of resin by processing about one liter of water, is possible

  1. Projecting LED product life based on application

    Science.gov (United States)

    Narendran, Nadarajah; Liu, Yi-wei; Mou, Xi; Thotagamuwa, Dinusha R.; Eshwarage, Oshadhi V. Madihe

    2016-09-01

    LED products have started to displace traditional light sources in many lighting applications. One of the commonly claimed benefits for LED lighting products is their long useful lifetime in applications. Today there are many replacement lamp products using LEDs in the marketplace. Typically, lifetime claims of these replacement lamps are in the 25,000-hour range. According to current industry practice, the time for the LED light output to reach the 70% value is estimated according to IESNA LM-80 and TM-21 procedures and the resulting value is reported as the whole system life. LED products generally experience different thermal environments and switching (on-off cycling) patterns when used in applications. Current industry test methods often do not produce accurate lifetime estimates for LED systems because only one component of the system, namely the LED, is tested under a continuous-on burning condition without switching on and off, and because they estimate for only one failure type, lumen depreciation. The objective of the study presented in this manuscript was to develop a test method that could help predict LED system life in any application by testing the whole LED system, including on-off power cycling with sufficient dwell time, and considering both failure types, catastrophic and parametric. The study results showed for the LED A-lamps tested in this study, both failure types, catastrophic and parametric, exist. The on-off cycling encourages catastrophic failure, and maximum operating temperature influences the lumen depreciation rate and parametric failure time. It was also clear that LED system life is negatively affected by on-off switching, contrary to commonly held belief. In addition, the study results showed that most of the LED systems failed catastrophically much ahead of the LED light output reaching the 70% value. This emphasizes the fact that life testing of LED systems must consider catastrophic failure in addition to lumen depreciation, and

  2. Configuration of LWR fuel enrichment or burnup yielding maximum power

    International Nuclear Information System (INIS)

    Bartosek, V.; Zalesky, K.

    1976-01-01

    An analysis is given of the spatial distribution of fuel burnup and enrichment in a light-water lattice of given dimensions with slightly enriched uranium, at which the maximum output is achieved. It is based on the spatial solution of neutron flux using a one-group diffusion model in which linear dependence may be expected of the fission cross section and the material buckling parameter on the fuel burnup and enrichment. Two problem constraints are considered, i.e., the neutron flux value and the specific output value. For the former the optimum core configuration remains qualitatively unchanged for any reflector thickness, for the latter the cases of a reactor with and without reflector must be distinguished. (Z.M.)

  3. Measurement of fission product release during LWR fuel failure

    International Nuclear Information System (INIS)

    Osetek, D.J.; King, J.J.

    1979-01-01

    The PBF is a specialized test reactor consisting of an annular core and a central test space 21 cm in diameter and 91 cm high. A test loop circulates coolant through the central experimental section at typical power reactor conditions. Light-water-reactor-type fuel rods are exposed to power bursts simulating reactivity insertion transients, and to power-cooling-mismatch conditions during which the rods are allowed to operate in film boiling. Fission product concentrations in the test loop coolant are continuously monitored during these transients by a Ge(Li) detector based gamma spectrometer. Automatic batch processing of pulse height spectra results in a list of radionuclide concentrations present in the loop coolant as a function of time during the test. Fission product behavior is then correlated to test parameters and posttest examination of the fuel rods. Data are presented from Test PCM-1

  4. Projected developments in the US and European LWR fabrication market

    International Nuclear Information System (INIS)

    Anderson, C.K.; Varley, G.

    1996-01-01

    There are several important factors influencing change in the fuel fabrication supply industry in Western Europe today. The most important is the changing supply relationships resulting from lower costs in the USA. With US fabrication prices substantially lower than in Western Europe, a situation which is sustainable based on production costs, a major force for change prevails which will have significant implications for the commercial development of both markets over the next decade. The extent to which supply from the USA will drive the Western Europe market will depend, to some extent, on three other factors: the new US-Euratom agreement on nuclear trade; electricity market deregulation in the European Union (EU); near term MOX fuel fabrication capacity. The main purpose of this paper is to discuss the current situation in the USA and Europe, and the manner in which this will lead to changing fuel fabrication supply dynamics in both markets. (author)

  5. LWR power plants lifetime management strategies. The Spanish approach

    International Nuclear Information System (INIS)

    Francia, Lorenzo; Carmena, Pio; Gomez Santamaria, Jesus; Regano, Martin; Marcelles, Ignacio

    2002-01-01

    During the last years, Spanish nuclear power plant (NPP) owners are increasingly concerned with optimising Plant Life Management. As a result of this interest, they are setting up Lifetime Management Programmes with the strategic objective to operate the NPPs beyond 40 years. In addition, a close link exists between life management work and improved maintenance practices concerning strategic, economic and safety aspects of NPP. As results of these important considerations, utilities will increase and globalise programmes for monitoring and mitigation of ageing. The afore mentioned programmes are based on knowledge of the precise condition of all components with the greatest effect on the economics and safety of the plant, and trends in their condition. The technical support for these programmes is: Methodologies and knowledge required identifying degradation mechanisms as a function of the characteristics of the components and operating conditions; Techniques for determining condition and trends; Analysis of the maintenance practices efficiency based on the above knowledge, techniques and methodologies; Improvement of maintenance practices for adequate mitigation and monitoring of ageing; Techniques and tools for collecting and ordering data about ageing and for condition assessment. Spanish NPPs have developed, under UNESA coordination, a standard methodology for lifetime management of LWRs. The methodology consists in the following main steps: Define with high accuracy the priorities and scope of the Life Management Programme. In this first step, plant systems, structures and components, important for lifetime management of LWRs, are selected and prioritised using a methodology based on screening criteria. The LCM Programme begins with an initial condition evaluation, which serves as the basis for establishing the main corrective and monitoring actions, and for preparing the first cost/benefit analyses for Life Management. The LCM Programme continues to progress

  6. LWR surveillance dosimetry improvement program: PSF metallurgical blind test results

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Maerker, R.E.; Stallmann, F.W.

    1984-01-01

    The metallurgical irradiation experiment at the Oak Ridge Research Reactor Poolside Facility (ORR-PSF) was designed as a benchmark to test the accuracy of radiation embrittlement predictions in the pressure vessel wall of light water reactors on the basis of results from surveillance capsules. The PSF metallurgical Blind Test is concerned with the simulated surveillance capsule (SSC) and the simulated pressure vessel capsule (SPVC). The data from the ORR-PSF benchmark experiment are the basis for comparison with the predictions made by participants of the metallurgical ''Blind Test''. The Blind Test required the participants to predict the embrittlement of the irradiated specimen based only on dosimetry and metallurgical data from the SSC1 capsule. This exercise included both the prediction of damage fluence and the prediction of embrittlement based on the predicted fluence. A variety of prediction methodologies was used by the participants. No glaring biases or other deficiencies were found, but neither were any of the methods clearly superior to the others. Closer analysis shows a rather complex and poorly understood relation between fluence and material damage. Many prediction formulas can give an adequate approximation, but further improvement of the prediction methodology is unlikely at this time given the many unknown factors. Instead, attention should be focused on determining realistic uncertainties for the predicted material changes. The Blind Test comparisons provide some clues for the size of these uncertainties. In particular, higher uncertainties must be assigned to materials whose chemical composition lies outside the data set for which the prediction formula was obtained. 16 references, 14 figures, 5 tables

  7. FORMATE-BASED FLUIDS: FORMULATION AND APPLICATION

    Directory of Open Access Journals (Sweden)

    Nediljka Gaurina-Međimurec

    2008-12-01

    Full Text Available Formate-based fluids has been successfully used in over hunders HPHT well operations since they introduced in field practice. They have many advantages when compared with conventional HPHT drilling and completion fluids such as: minimal formation damage, maintenance of additve properties at high temperatures, reduced hydraulic flow resistance, low potential for differential sticking, naturally lubricating, very low corrosion rates, biodegradable and pose little risk to the environment etc. Formate-based fluids can be applied during deep slim hole drilling, shale drilling, reservoir drilling, salt and gas hydrate formations drilling. The laboratory research was carried out to evaluate the rheological behavior of formate-based fluids as a function of temperature. Formate-based fluids were formulated using potassium formate brine, xanthan polymer, PAC, starch and calcium carbonate. Experimental results show that potassium formate improves the thermal stability of polymers.

  8. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented

  9. Prediction of fission product and aerosol behaviour during a postulated severe accident in a LWR

    International Nuclear Information System (INIS)

    Guentay, S.; Aeby, F.; Raguin, M.; Passalacqua, R.

    1990-02-01

    Lack of appropriate energy removal causes fuel elements in a reactor core to overheat and may eventually cause core to degrade. Fission products will be emitted from a degraded reactor core. Aerosols are generated when the vapours of various fuel and structural materials reach a cold environment and nucleate. In addition to the fission products release and aerosol generation taking place in the reactor vessel, some more fission products release and aerosol generation will occur when the molten core debris leaves the pressure vessel bottom head and comes in contact with the pedestal concrete floor. Fission products, if they are released to environment from the containment boundary, exert a great danger to public health. A source term is defined as the quantity, timing, and characteristics of the release of radionuclide material to the environment following a postulated severe accident. At PSI a considerable effort hase been spent in investigating and establishing a source term assessment methodology in order to predict the source term for a given Light Water Reactor (LWR) accident scenario. This report introduces the computer programs and the methods associated with the release of the fission products, generation of the aerosols and behaviour of the aerosols in LWR compartments used for a source term assessment analysis at PSI. (author) 4 figs., 5 tabs., 28 refs

  10. In-core materials testing under LWR conditions in the Halden reactor

    International Nuclear Information System (INIS)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

    2002-01-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  11. Data needs for long-term dry storage of LWR fuel. Interim report

    International Nuclear Information System (INIS)

    Einziger, R.E.; Baldwin, D.L.; Pitman, S.G.

    1998-04-01

    The NRC approved dry storage of spent fuel in an inert environment for a period of 20 years pursuant to 10CFR72. However, at-reactor dry storage of spent LWR fuel may need to be implemented for periods of time significantly longer than the NRC's original 20-year license period, largely due to uncertainty as to the date the US DOE will begin accepting commercial spent fuel. This factor is leading utilities to plan not only for life-of-plant spent-fuel storage during reactor operation but also for the contingency of a lengthy post-shutdown storage. To meet NRC standards, dry storage must (1) maintain subcriticality, (2) prevent release of radioactive material above acceptable limits, (3) ensure that radiation rates and doses do not exceed acceptable limits, and (4) maintain retrievability of the stored radioactive material. In light of these requirements, this study evaluates the potential for storing spent LWR fuel for up to 100 years. It also identifies major uncertainties as well as the data required to eliminate them. Results show that the lower radiation fields and temperatures after 20 years of dry storage promote acceptable fuel behavior and the extension of storage for up to 100 years. Potential changes in the properties of dry storage system components, other than spent-fuel assemblies, must still be evaluated

  12. Decay heat and gamma dose-rate prediction capability in spent LWR fuel

    International Nuclear Information System (INIS)

    Neely, G.J.; Schmittroth, F.

    1982-08-01

    The ORIGEN2 code was established as a valid means to predict decay heat from LWR spent fuel assemblies for decay times up to 10,000 year. Calculational uncertainties ranged from 8.6% to a maximum of 16% at 2.5 years and 300 years cooling time, respectively. The calculational uncertainties at 2.5 years cooling time are supported by experiment. Major sources of uncertainty at the 2.5 year cooling time were identifed as irradiation history (5.7%) and nuclear data together with calculational methods (6.3%). The QAD shielding code was established as a valid means to predict interior and exterior gamma dose rates of spent LWR fuel assemblies. A calculational/measurement comparison was done on two assemblies with different irradiation histories and supports a 35% calculational uncertainty at the 1.8 and 3.0 year decay times studied. Uncertainties at longer times are expected to increase, but not significantly, due to an increased contribution from the actinides whose inventories are assigned a higher uncertainty. The uncertainty in decay heat rises to a maximum of 16% due to actinide uncertainties. A previous study was made of the neutron emission rate from a typical Turkey Point Unit 3, Region 4 spent fuel assembly at 5 years decay time. A conservative estimate of the neutron dose rate at the assembly surface was less than 0.5 rem/hr

  13. On the Diffusion Coefficient of Two-step Method for LWR analysis

    International Nuclear Information System (INIS)

    Lee, Deokjung; Choi, Sooyoung; Smith, Kord S.

    2015-01-01

    The few-group constants including diffusion coefficients are generated from the assembly calculation results. Once the assembly calculation is done, the cross sections (XSs) are spatially homogenized, and a critical spectrum calculation is performed in order to take into account the neutron leakages of the lattice. The diffusion coefficient is also generated through the critical spectrum calculation. Three different methods of the critical spectrum calculation such as B1 method, P1 method, and fundamental mode (FM) calculation method are considered in this paper. The diffusion coefficients can also be affected by transport approximations for the transport XS calculation which is used in the assembly transport lattice calculation in order to account for the anisotropic scattering effects. The outflow transport approximation and the inflow transport approximation are investigated in this paper. The accuracy of the few group data especially the diffusion coefficients has been studied to optimize the combination of the transport correction methods and the critical spectrum calculation methods using the UNIST lattice physics code STREAM. The combination of the inflow transport approximation and the FM method is shown to provide the highest accuracy in the LWR core calculations. The methodologies to calculate the diffusion coefficients have been reviewed, and the performances of them have been investigated with a LWR core problem. The combination of the inflow transport approximation and the fundamental mode critical spectrum calculation shows the smallest errors in terms of assembly power distribution

  14. Long-term embrittlement of cast duplex stainless steels in LWR systems

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.

    1990-08-01

    This progress report summarizes work performed by Argonne National Laboratory on long-term embrittlement of cast duplex stainless steels in LWR systems during the six months from April to September 1988. Characteristics of the primary mechanism of aging embrittlement (i.e., spinodal decomposition of ferrite) and synergistic effects of alloying and impurity elements that influence the kinetics of the primary mechanism are discussed. Several secondary metallurgical processes of embrittlement, strongly dependent on the C, N, Ni, Mo, and Si content of various heats, are identified. Information on kinetics and data on impact properties are analyzed and correlated with microstructural characteristics to provide a unified method of extrapolating accelerated-aging data to reactor operating conditions. Fracture toughness data are presented for several heats of cast stainless steel aged at temperatures between 320 and 450 degrees C for times up to 10,000 h. Mechanical property data are analyzed to develop the procedure and correlations or predicting the kinetics and extent of embrittlement of reactor components from known material parameters. The method and examples of estimating the impact strength and fracture toughness of cast components during reactor service are described. The lower-bound values of impact strength and fracture toughness for cast stainless steels at LWR operating temperatures are defined. 42 refs., 14 figs., 6 tabs

  15. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    International Nuclear Information System (INIS)

    Schanz, G.; Hagen, S.; Hofmann, P.; Sepold, L.; Schumacher, G.

    1992-01-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400deg C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4 C), which are leading to extensive low-temperature melt formation around 1200deg C. Interrelations between those basic phenomena, resulting for example in cladding deformation ('flowering') and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ('quenching') are determining the evolution paths of fuel element destruction, which are to be identified. (orig.)

  16. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  17. The Width of High Burnup Structure in LWR UO2 Fuel

    International Nuclear Information System (INIS)

    Koo, Yang-Hyun; Lee, Byung-Ho; Oh, Jae-Yong; Sohn, Dong-Seong

    2007-01-01

    The measured data available in the open literature on the width of high burnup structure (HBS) in LWR UO 2 fuel were analyzed in terms of pellet average burnup, enrichment, and grain size. Dependence of the HBS width on pellet average burnup was shown to be divided into three regions; while the HBS width is governed by accumulation of fission damage (i.e., burnup) for burnup below 60 GWd/tU, it seems to be restricted to some limiting value of around 1.5 mm for burnup above 75 GWd/tU due to high temperature which might have caused extensive annealing of irradiation damage. As for intermediate burnup between 60 and 75 GWd/tU, although temperature would not have been so high as to induce extensive annealing, the microstructural damage could have been partly annealed, resulting in the reduction of the HBS width. It was found that both enrichment and grain size also affects the HBS width. However, as long as the pellet average burnup is lower than about 75 GWd/tU, the effect does not appear to be significant for the enrichment and grain size that are typically used in current LWR fuel. (authors)

  18. MCOR - Monte Carlo depletion code for reference LWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Puente Espel, Federico, E-mail: fup104@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Tippayakul, Chanatip, E-mail: cut110@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Ivanov, Kostadin, E-mail: kni1@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University (United States); Misu, Stefan, E-mail: Stefan.Misu@areva.com [AREVA, AREVA NP GmbH, Erlangen (Germany)

    2011-04-15

    Research highlights: > Introduction of a reference Monte Carlo based depletion code with extended capabilities. > Verification and validation results for MCOR. > Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations. Additionally

  19. MCOR - Monte Carlo depletion code for reference LWR calculations

    International Nuclear Information System (INIS)

    Puente Espel, Federico; Tippayakul, Chanatip; Ivanov, Kostadin; Misu, Stefan

    2011-01-01

    Research highlights: → Introduction of a reference Monte Carlo based depletion code with extended capabilities. → Verification and validation results for MCOR. → Utilization of MCOR for benchmarking deterministic lattice physics (spectral) codes. - Abstract: The MCOR (MCnp-kORigen) code system is a Monte Carlo based depletion system for reference fuel assembly and core calculations. The MCOR code is designed as an interfacing code that provides depletion capability to the LANL Monte Carlo code by coupling two codes: MCNP5 with the AREVA NP depletion code, KORIGEN. The physical quality of both codes is unchanged. The MCOR code system has been maintained and continuously enhanced since it was initially developed and validated. The verification of the coupling was made by evaluating the MCOR code against similar sophisticated code systems like MONTEBURNS, OCTOPUS and TRIPOLI-PEPIN. After its validation, the MCOR code has been further improved with important features. The MCOR code presents several valuable capabilities such as: (a) a predictor-corrector depletion algorithm, (b) utilization of KORIGEN as the depletion module, (c) individual depletion calculation of each burnup zone (no burnup zone grouping is required, which is particularly important for the modeling of gadolinium rings), and (d) on-line burnup cross-section generation by the Monte Carlo calculation for 88 isotopes and usage of the KORIGEN libraries for PWR and BWR typical spectra for the remaining isotopes. Besides the just mentioned capabilities, the MCOR code newest enhancements focus on the possibility of executing the MCNP5 calculation in sequential or parallel mode, a user-friendly automatic re-start capability, a modification of the burnup step size evaluation, and a post-processor and test-matrix, just to name the most important. The article describes the capabilities of the MCOR code system; from its design and development to its latest improvements and further ameliorations

  20. Development of methods for burn-up calculations for LWR's

    International Nuclear Information System (INIS)

    Jaschik, W.

    1978-01-01

    This method is based on all burn-up depending data, namely particle densities and neutron spectra, being available in a burn-up library. This one is created by means of a small number of cell burn-up calculations which can easily be carried out and in which the heterogeneous cell structure and self-shielding effects can explicitly be accounted for. Then the cluster burn-up is simulated by adequate correlation of the burn-up data. The advantage of this method is given by - an exact determination of the real spectrum distribution in the individual fuel element clusters; - an exact determination of the burn-up related spectrum variations for each fuel rod and for each burn-up value obtained; - accounting for heterogeneity of the fuel rod cells and the self-shielding in the fuel; high accuracy of the results of a comparably low effort and - simple handling by largely automating the process of computation. Programed realization was achieved by establishing the RSYST modules ABRAJA, MITHOM, and SIMABB and their implementation within the code system. (orig./HP) [de

  1. Fast reactors will eat nuclear waste from LWR

    International Nuclear Information System (INIS)

    Tokiwai, Moriyasu

    1999-01-01

    Although nuclear power is one of the indispensable energy sources to support modern life styles in developed countries, it becomes harder and harder to increase its capacity. Newspaper reported that there are numbers of evidences showing the suppression effect on cancer by the low level of radiation. It is expected for public people that the fear for radiation induced harm on health will mitigate through the explanation based on scientific evidences. Safe management of radioactive waste is one of the most serious issues to be solved. The neutron at fast reactors can eat more effectively the long lived several nuclear waste materials from light water reactor system, The key issue is to develop the fast reactor fuel cycle system technologies that are more economical, more proliferation resistant and higher breeding ratio. The Metallic Fuel Cycle is one of the options for the future fast breeder reactor and its related fuel cycle that enable to give the answer for the radioactive waste issues. The attractiveness of the metallic fuel cycle concept is briefly described. (author)

  2. Survey of European LWR fuel irradiation test facilities

    Energy Technology Data Exchange (ETDEWEB)

    Hardt, P von der [Commission of the European Communities, Joint Research Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    The first European commercial nuclear power plants (1956) featured gas-cooled thermal reactors. Although there is now a general orientation towards light water cooled plants (with a slight preference for the PWR) a large fraction of the 1982 nuclear generating capacity is still invested in gas-cooled reactors. R and D also continues for the HTGR with its long-term development potential. This paper, however, is limited to a general survey of experimental programmes and facilities for light water reactor fuel testing in Western Europe, particularly inside the European Communities. As it turns out, over a dozen major installations are available, all connected to research reactors in government-funded R and D centres. Their equipment is briefly reviewed. Some 50% of the experimental programmes are carried out in large international collaboration, involving up to 20 organizations per project. Techniques and results are rapidly communicated through frequent meetings and conferences. It is anticipated that a part of the present research reactor-based work will gradually shift to power reactor pool side inspection facilities. (author)

  3. Management feature of transuranic for HTGR and LWR

    International Nuclear Information System (INIS)

    Wei Jinfeng; Li Fu; Sun Yuliang

    2013-01-01

    Long-lived actinides from spent fuels can cause potential long-term environ- mental hazards. The generation and incineration of transuranic in different closed fuel cycles were studied. U and Pu were recycled from spent fuel in the 250 MW high-temperature gas-cooled reactor-pebble-bed-module (HTR-PM) U-Pu fuelled core, and then PuO 2 and MOX fuel elements were designed based on this recycled U and Pu. These fuel elements were used to build up a new PuO 2 or MOX fuelled core with the same geometry of the original reactor. Characteristics of transuranic incineration with HTGR open and closed fuel cycles were studied with VSOP code, and the corresponding results from the light water reactor were compared and analyzed. The transuranic generation with HTGR open fuel cycle is almost half of the corresponding result of the light water reactor. Thus, HTGR closed fuel cycles can effectively burn transuranic. (authors)

  4. Fast reactors will eat nuclear waste from LWR

    Energy Technology Data Exchange (ETDEWEB)

    Tokiwai, Moriyasu [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Res., Lab.

    1999-12-01

    Although nuclear power is one of the indispensable energy sources to support modern life styles in developed countries, it becomes harder and harder to increase its capacity. Newspaper reported that there are numbers of evidences showing the suppression effect on cancer by the low level of radiation. It is expected for public people that the fear for radiation induced harm on health will mitigate through the explanation based on scientific evidences. Safe management of radioactive waste is one of the most serious issues to be solved. The neutron at fast reactors can eat more effectively the long lived several nuclear waste materials from light water reactor system, The key issue is to develop the fast reactor fuel cycle system technologies that are more economical, more proliferation resistant and higher breeding ratio. The Metallic Fuel Cycle is one of the options for the future fast breeder reactor and its related fuel cycle that enable to give the answer for the radioactive waste issues. The attractiveness of the metallic fuel cycle concept is briefly described. (author)

  5. Survey of European LWR fuel irradiation test facilities

    International Nuclear Information System (INIS)

    Hardt, P. von der

    1983-01-01

    The first European commercial nuclear power plants (1956) featured gas-cooled thermal reactors. Although there is now a general orientation towards light water cooled plants (with a slight preference for the PWR) a large fraction of the 1982 nuclear generating capacity is still invested in gas-cooled reactors. R and D also continues for the HTGR with its long-term development potential. This paper, however, is limited to a general survey of experimental programmes and facilities for light water reactor fuel testing in Western Europe, particularly inside the European Communities. As it turns out, over a dozen major installations are available, all connected to research reactors in government-funded R and D centres. Their equipment is briefly reviewed. Some 50% of the experimental programmes are carried out in large international collaboration, involving up to 20 organizations per project. Techniques and results are rapidly communicated through frequent meetings and conferences. It is anticipated that a part of the present research reactor-based work will gradually shift to power reactor pool side inspection facilities. (author)

  6. LWR containment thermal hydraulic codes benchmark demona B3 exercise

    International Nuclear Information System (INIS)

    Della Loggia, E.; Gauvain, J.

    1988-01-01

    Recent discussion about the aerosol codes currently used for the analysis of containment retention capabilities have revealed a number of questions concerning the reliabilities and verifications of the thermal-hydraulic modules of these codes with respect to the validity of implemented physical models and the stability and effectiveness of numerical schemes. Since these codes are used for the calculation of the Source Term for the assessment of radiological consequences of severe accidents, they are an important part of reactor safety evaluation. For this reason the Commission of European Communities (CEC), following the recommendation mode by experts from Member Stades, is promoting research in this field with the aim also of establishing and increasing collaboration among Research Organisations of member countries. In view of the results of the studies, the CEC has decided to carry out a Benchmark exercise for severe accident containment thermal hydraulics codes. This exercise is based on experiment B3 in the DEMONA programme. The main objective of the benchmark exercise has been to assess the ability of the participating codes to predict atmosphere saturation levels and bulk condensation rates under conditions similar to those predicted to follow a severe accident in a PWR. This exercise follows logically on from the LA-4 exercise, which, is related to an experiment with a simpler internal geometry. We present here the results obtained so far and from them preliminary conclusions are drawn, concerning condensation temperature, pressure, flow rates, in the reactor containment

  7. Summary of LWR safety research in the USA

    International Nuclear Information System (INIS)

    Murley, T.E.; Tong, L.S.; Bennett, G.L.

    1977-01-01

    quantification of reactor behaviour under off-normal conditions. The reactor operational safety research is just starting initially, it addresses fire protection, component aging and human engineering. To date, the NRC water reactor safety research program has greatly expanded the safety data base which in turn aids the further quantification of the inherent safety margins in nuclear power plants

  8. Polymer Based Biosensors for Medical Applications

    DEFF Research Database (Denmark)

    Cherré, Solène; Rozlosnik, Noemi

    2015-01-01

    , environmental monitoring and food safety. The detected element varies from a single molecule (such as glucose), a biopolymer (such as DNA or a protein) to a whole organism (such as bacteria). Due to their easy use and possible miniaturization, biosensors have a high potential to come out of the lab...... and be available for use by everybody. To fulfil these purposes, polymers represent very appropriate materials. Many nano- and microfabrication methods for polymers are available, allowing a fast and cheap production of devices. This chapter will present the general concept of a biosensor in a first part......The objective of this chapter is to give an overview about the newest developments in biosensors made of polymers for medical applications. Biosensors are devices that can recognize and detect a target with high selectivity. They are widely used in many fields such as medical diagnostic...

  9. Implementation of ion based applications for safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, R.L.; Johnson, C.S.

    1996-08-01

    The uses of the LON, `Local Operating Network,` developed by Echelon Corporation, Palo Alto, California, has been expanded to handle a number of safeguards applications. A magnetic and vibration sensor pack has been developed to monitor for magnetic fields and vibration. This sensor pack can be attached to any source that generates a magnetic field, such as electrical solenoids or motors, to detect when the source is activated. New network nodes that interface directly with the raw data of Sandia developed radiation sensors, for detecting the presence of radiation sources, have been developed. The capacity of the network has been expanded to allow the transmission of large data sets, specifically the transmission of digital video images from the Sandia-developed-Image Compression and Authentication Module (ICAM).

  10. Indium antimonide based HEMT for RF applications

    International Nuclear Information System (INIS)

    Subash, T. D.; Gnanasekaran, T.

    2014-01-01

    We report on an indium antimonide high electron mobility transistor with record cut-off frequency characteristics. For high frequency response it is important to minimize parasitic resistance and capacitance to improve short-channel effects. For analog applications adequate pinch-off behavior is demonstrated. For proper device scaling we need high electron mobility and high electron density. Toward this end, the device design features and simulation are carried out by the Synopsys TCAD tool. A 30 nm InSb HEMT exhibits an excellent cut-off frequency of 586 GHz. To the knowledge of the authors, the obtained cut-off frequency is the highest ever reported in any FET on any material system. (semiconductor materials)

  11. Gamifying Navigation in Location-Based Applications

    DEFF Research Database (Denmark)

    Nadarajah, Stephanie Githa; Overgaard, Benjamin Nicholas; Pedersen, Peder Walz

    2017-01-01

    Location-based games entertain players usually by interactions at points of interest (POIs). Navigation between POIs often involve the use of either a physical or digital map, not taking advantage of the opportunity available to engage users in activities between POIs. The paper presents riddle s...

  12. Risk-based remediation: Approach and application

    International Nuclear Information System (INIS)

    Frishmuth, R.A.; Benson, L.A.

    1995-01-01

    The principle objective of remedial actions is to protect human health and the environment. Risk assessments are the only defensible tools available to demonstrate to the regulatory community and public that this objective can be achieved. Understanding the actual risks posed by site-related contamination is crucial to designing cost-effective remedial strategies. All to often remedial actions are overdesigned, resulting in little to no increase in risk reduction while increasing project cost. Risk-based remedial actions have recently been embraced by federal and state regulators, industry, government, the scientific community, and the public as a mechanism to implement rapid and cost-effective remedial actions. Emphasizing risk reduction, rather than adherence to ambiguous and generic standards, ensures that only remedial actions required to protect human health and the environment at a particular site are implemented. Two sites are presented as case studies on how risk-based approaches are being used to remediate two petroleum hydrocarbon contaminated sites. The sites are located at two US Air Force Bases, Wurtsmith Air Force Base (AFB) in Oscoda, Michigan and Malmstrom AFB in Great Falls, Montana

  13. A new water-based liquid scintillator and potential applications

    Energy Technology Data Exchange (ETDEWEB)

    Yeh, M., E-mail: yeh@bnl.gov [Chemistry Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Hans, S.; Beriguete, W.; Rosero, R.; Hu, L.; Hahn, R.L. [Chemistry Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Diwan, M.V.; Jaffe, D.E.; Kettell, S.H.; Littenberg, L. [Physics Department, Brookhaven National Laboratory, Upton, NY 11973 (United States)

    2011-12-21

    In this paper we describe a new type of scintillating liquid based on water. We describe the concept, preparation, and properties of this liquid, and how it could be used for a very large, but economical detector. The applications of such a detector range from fundamental physics such as nucleon decay and neutrino physics to physics with broader application such as neutron detection. We briefly describe the scientific requirements of these applications, and how they can be satisfied by the new material.

  14. Biometrics and smart card based applications for nuclear industry

    International Nuclear Information System (INIS)

    Nishanth Reddy, J.; Dheeraj Reddy, J.; Narender Reddy, J.

    2004-01-01

    Biometrics has emerged as a convenient, foolproof and well-accepted technology for identification around the globe. Nucleonix has developed innovative solutions based on finger scan biometrics for various industries. This paper closely looks into the application areas for the nuclear industry and how it will benefit this industry, in terms of identification, access control, security of PCs and applications, attendance, machinery usage control and other custom applications. (author)

  15. Mobile Application Identification based on Hidden Markov Model

    Directory of Open Access Journals (Sweden)

    Yang Xinyan

    2018-01-01

    Full Text Available With the increasing number of mobile applications, there has more challenging network management tasks to resolve. Users also face security issues of the mobile Internet application when enjoying the mobile network resources. Identifying applications that correspond to network traffic can help network operators effectively perform network management. The existing mobile application recognition technology presents new challenges in extensibility and applications with encryption protocols. For the existing mobile application recognition technology, there are two problems, they can not recognize the application which using the encryption protocol and their scalability is poor. In this paper, a mobile application identification method based on Hidden Markov Model(HMM is proposed to extract the defined statistical characteristics from different network flows generated when each application starting. According to the time information of different network flows to get the corresponding time series, and then for each application to be identified separately to establish the corresponding HMM model. Then, we use 10 common applications to test the method proposed in this paper. The test results show that the mobile application recognition method proposed in this paper has a high accuracy and good generalization ability.

  16. DNA-Based Applications in Nanobiotechnology

    Directory of Open Access Journals (Sweden)

    Khalid M. Abu-Salah

    2010-01-01

    Full Text Available Biological molecules such as deoxyribonucleic acid (DNA have shown great potential in fabrication and construction of nanostructures and devices. The very properties that make DNA so effective as genetic material also make it a very suitable molecule for programmed self-assembly. The use of DNA to assemble metals or semiconducting particles has been extended to construct metallic nanowires and functionalized nanotubes. This paper highlights some important aspects of conjugating the unique physical properties of dots or wires with the remarkable recognition capabilities of DNA which could lead to miniaturizing biological electronics and optical devices, including biosensors and probes. Attempts to use DNA-based nanocarriers for gene delivery are discussed. In addition, the ecological advantages and risks of nanotechnology including DNA-based nanobiotechnology are evaluated.

  17. Void reactivity feedback analysis for U-based and Th-based LWR incineration cycles

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, B.A.; Parks, G.T. [Cambridge University Engineering Department, Trumpington Street, Cambridge, CB2 1PZ (United Kingdom); Franceschini, F. [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    In reduced-moderation LWRs, an external supply of transuranic (TRU) can be incinerated by mixing it with a fertile isotope ({sup 238}U or {sup 232}Th) and recycling all the actinides after each cycle. Performance is limited by coolant reactivity feedback - the moderator density coefficient (MDC) must be kept negative. The MDC is worse when more TRU is loaded, but TRU feed is also needed to maintain criticality. To assess the performance of this fuel cycle in different neutron spectra, three LWRs are considered: 'reference' PWRs and reduced-moderation PWRs and BWRs. The MDC of the equilibrium cycle is analysed by reactivity decomposition with perturbed coolant density by isotope and neutron energy. The results show that using {sup 232}Th as a fertile isotope yields superior performance to {sup 238}U. This is due essentially to the high resonance η of U bred from Th (U3), which increases the fissility of the U3-TRU isotope vector in the Th-fueled system relative to the U-fueled system, and also improves the MDC in a sufficiently hard spectrum. Spatial separation of TRU and U3 in the Th-fueled system renders further improvement by hardening the neutron spectrum in the TRU and softening it in the U3. This improves the TRU η and increases the negative MDC contribution from reduced thermal fission in U3. (authors)

  18. Invariant-Based Automatic Testing of Modern Web Applications

    NARCIS (Netherlands)

    Mesbah, A.; Van Deursen, A.; Roest, D.

    2011-01-01

    AJAX-based Web 2.0 applications rely on stateful asynchronous client/server communication, and client-side run-time manipulation of the DOM tree. This not only makes them fundamentally different from traditional web applications, but also more error-prone and harder to test. We propose a method for

  19. Development of Web-Based Learning Application for Generation Z

    Science.gov (United States)

    Hariadi, Bambang; Dewiyani Sunarto, M. J.; Sudarmaningtyas, Pantjawati

    2016-01-01

    This study aimed to develop a web-based learning application as a form of learning revolution. The form of learning revolution includes the provision of unlimited teaching materials, real time class organization, and is not limited by time or place. The implementation of this application is in the form of hybrid learning by using Google Apps for…

  20. NUCLEAR SAFETY DESIGN BASES FOR LICENSE APPLICATION

    International Nuclear Information System (INIS)

    Garrett, R.J.

    2005-01-01

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111 [DIRS 156605] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113 [DIRS 156605] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

  1. Neural Network Based Models for Fusion Applications

    Science.gov (United States)

    Meneghini, Orso; Tema Biwole, Arsene; Luda, Teobaldo; Zywicki, Bailey; Rea, Cristina; Smith, Sterling; Snyder, Phil; Belli, Emily; Staebler, Gary; Canty, Jeff

    2017-10-01

    Whole device modeling, engineering design, experimental planning and control applications demand models that are simultaneously physically accurate and fast. This poster reports on the ongoing effort towards the development and validation of a series of models that leverage neural-­network (NN) multidimensional regression techniques to accelerate some of the most mission critical first principle models for the fusion community, such as: the EPED workflow for prediction of the H-Mode and Super H-Mode pedestal structure the TGLF and NEO models for the prediction of the turbulent and neoclassical particle, energy and momentum fluxes; and the NEO model for the drift-kinetic solution of the bootstrap current. We also applied NNs on DIII-D experimental data for disruption prediction and quantifying the effect of RMPs on the pedestal and ELMs. All of these projects were supported by the infrastructure provided by the OMFIT integrated modeling framework. Work supported by US DOE under DE-SC0012656, DE-FG02-95ER54309, DE-FC02-04ER54698.

  2. Nuclear Safety Design Base for License Application

    International Nuclear Information System (INIS)

    R.J. Garrett

    2005-01-01

    The purpose of this report is to identify and document the nuclear safety design requirements that are specific to structures, systems, and components (SSCs) of the repository that are important to safety (ITS) during the preclosure period and to support the preclosure safety analysis and the license application for the high-level radioactive waste (HLW) repository at Yucca Mountain, Nevada. The scope of this report includes the assignment of nuclear safety design requirements to SSCs that are ITS and does not include the assignment of design requirements to SSCs or natural or engineered barriers that are important to waste isolation (ITWI). These requirements are used as input for the design of the SSCs that are ITS such that the preclosure performance objectives of 10 CFR 63.111(b) [DIRS 173273] are met. The natural or engineered barriers that are important to meeting the postclosure performance objectives of 10 CFR 63.113(b) and (c) [DIRS 173273] are identified as ITWI. Although a structure, system, or component (SSC) that is ITS may also be ITWI, this report is only concerned with providing the nuclear safety requirements for SSCs that are ITS to prevent or mitigate event sequences during the repository preclosure period

  3. Investigation of RFID Based Sensors for Sustainable Transportation Applications

    Science.gov (United States)

    2011-01-21

    Through support of a University Transportation Research Center Faculty Development Minigrant an investigation was made into the use of RFID based sensing technologies for transportation purposes. Transportation applications would potentially include ...

  4. The Component-Based Application for GAMESS

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Fang [Iowa State Univ., Ames, IA (United States)

    2007-01-01

    GAMESS, a quantum chetnistry program for electronic structure calculations, has been freely shared by high-performance application scientists for over twenty years. It provides a rich set of functionalities and can be run on a variety of parallel platforms through a distributed data interface. While a chemistry computation is sophisticated and hard to develop, the resource sharing among different chemistry packages will accelerate the development of new computations and encourage the cooperation of scientists from universities and laboratories. Common Component Architecture (CCA) offers an enviromnent that allows scientific packages to dynamically interact with each other through components, which enable dynamic coupling of GAMESS with other chetnistry packages, such as MPQC and NWChem. Conceptually, a cotnputation can be constructed with "plug-and-play" components from scientific packages and require more than componentizing functions/subroutines of interest, especially for large-scale scientific packages with a long development history. In this research, we present our efforts to construct cotnponents for GAMESS that conform to the CCA specification. The goal is to enable the fine-grained interoperability between three quantum chemistry programs, GAMESS, MPQC and NWChem, via components. We focus on one of the three packages, GAMESS; delineate the structure of GAMESS computations, followed by our approaches to its component development. Then we use GAMESS as the driver to interoperate integral components from the other tw"o packages, arid show the solutions for interoperability problems along with preliminary results. To justify the versatility of the design, the Tuning and Analysis Utility (TAU) components have been coupled with GAMESS and its components, so that the performance of GAMESS and its components may be analyzed for a wide range of systetn parameters.

  5. Technology-based suicide prevention: current applications and future directions.

    Science.gov (United States)

    Luxton, David D; June, Jennifer D; Kinn, Julie T

    2011-01-01

    This review reports on current and emerging technologies for suicide prevention. Technology-based programs discussed include interactive educational and social networking Web sites, e-mail outreach, and programs that use mobile devices and texting. We describe innovative applications such as virtual worlds, gaming, and text analysis that are currently being developed and applied to suicide prevention and outreach programs. We also discuss the benefits and limitations of technology-based applications and discuss future directions for their use.

  6. Electrical conduction in solid materials physicochemical bases and possible applications

    CERN Document Server

    Suchet, J P

    2013-01-01

    Electrical Conduction in Solid Materials (Physicochemical Bases and Possible Applications) investigates the physicochemical bases and possible applications of electrical conduction in solid materials, with emphasis on conductors, semiconductors, and insulators. Topics range from the interatomic bonds of conductors to the effective atomic charge in conventional semiconductors and magnetic transitions in switching semiconductors. Comprised of 10 chapters, this volume begins with a description of electrical conduction in conductors and semiconductors, metals and alloys, as well as interatomic bon

  7. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  8. FPGA based accelerators for financial applications

    CERN Document Server

    2015-01-01

    This book covers the latest approaches and results from reconfigurable computing architectures employed in the finance domain. So-called field-programmable gate arrays (FPGAs) have already shown to outperform standard CPU- and GPU-based computing architectures by far, saving up to 99% of energy depending on the compute tasks. Renowned authors from financial mathematics, computer architecture, and finance business introduce the readers into today’s challenges in finance IT, illustrate the most advanced approaches and use cases, and present currently known methodologies for integrating FPGAs in finance systems together with latest results. The complete algorithm-to-hardware flow is covered holistically, so this book serves as a hands-on guide for IT managers, researchers, and quants/programmers who think about integrating FPGAs into their current IT systems.

  9. New Developments of Ti-Based Alloys for Biomedical Applications

    Science.gov (United States)

    Li, Yuhua; Yang, Chao; Zhao, Haidong; Qu, Shengguan; Li, Xiaoqiang; Li, Yuanyuan

    2014-01-01

    Ti-based alloys are finding ever-increasing applications in biomaterials due to their excellent mechanical, physical and biological performance. Nowdays, low modulus β-type Ti-based alloys are still being developed. Meanwhile, porous Ti-based alloys are being developed as an alternative orthopedic implant material, as they can provide good biological fixation through bone tissue ingrowth into the porous network. This paper focuses on recent developments of biomedical Ti-based alloys. It can be divided into four main sections. The first section focuses on the fundamental requirements titanium biomaterial should fulfill and its market and application prospects. This section is followed by discussing basic phases, alloying elements and mechanical properties of low modulus β-type Ti-based alloys. Thermal treatment, grain size, texture and properties in Ti-based alloys and their limitations are dicussed in the third section. Finally, the fourth section reviews the influence of microstructural configurations on mechanical properties of porous Ti-based alloys and all known methods for fabricating porous Ti-based alloys. This section also reviews prospects and challenges of porous Ti-based alloys, emphasizing their current status, future opportunities and obstacles for expanded applications. Overall, efforts have been made to reveal the latest scenario of bulk and porous Ti-based materials for biomedical applications. PMID:28788539

  10. Activity- Based Costing: An Application In The Olive Sector

    OpenAIRE

    Bekçioğlu, Selim; Gürel, Eymen; Kızılyalçın, D. Ali

    2014-01-01

    The purpose of the study named as “The Activity Based Costing: An Application in the Olive Sector” is the implementation of activity based costing in olive sector and the statement of the differences as regards to the traditional costing. The accepted basic hypothesis in the thesis is that the activity-based costing provides more accurate cost information than the traditional costing process. In this study initially comprehensive information about activity based costing is given and after th...

  11. LWR high burn-up operation and MOX introduction. Fuel cycle performance from the viewpoint of waste management

    International Nuclear Information System (INIS)

    Inagaki, Yaohiro; Iwasaki, Tomohiko; Niibori, Yuichi; Sato, Seichi; Ohe, Toshiaki; Kato, Kazuyuki; Torikai, Seishi; Nagasaki, Shinya; Kitayama, Kazumi

    2009-01-01

    From the viewpoint of waste management, a quantitative evaluation of LWR nuclear fuel cycle system performance was carried out, considering both higher burn-up operation of UO 2 fuel coupled with the introduction of MOX fuel. A major parameter to quantify this performance is the number of high-level waste (HLW) glass units generated per GWd (gigawatt-day based on reactor thermal power generation before electrical conversion). This parameter was evaluated for each system up to a maximum burn-up of 70GWd/THM (gigawatt-day per ton of heavy metal) assuming current conventional reprocessing and vitrification conditions where the waste loading of glass is restricted by the heat generation rate, the MoO 3 content, or the noble metal content. The results showed that higher burn-up operation has no significant influence on the number of glass units generated per GWd for UO 2 fuel, though the number of glass units per THM increases linearly with burn-up and is restricted by the heat generation rate. On the other hand, the introduction of MOX fuel causes the number of glass units per GWd to double owing to the increase in the heat generation rate. An extended cooling period of the spent fuel prior to reprocessing effectively reduces the heat generation rate for UO 2 fuel, while a separation of minor actinides (Np, Am, and Cm) from the high-level waste provides additional reduction for MOX fuel. However, neither of these leads to a substantial reduction in the number of glass units, since the MoO 3 content or the noble metal content restricts the number of glass units rather than the heat generation rate. These results suggest that both the MoO 3 content and the noble metal content provide the key to reducing the amount of waste glass that is generated, leading to an overall improvement in fuel cycle system performance. (author)

  12. Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

    International Nuclear Information System (INIS)

    Loesoenen, Pekka

    2002-01-01

    A model for the release of stable fission gases by diffuion from sintered LWR UO 2 fuel grains is presented. The model takes into account intragranular gas bubble behaviour as a function of grain radius. The bubbles are assumed to be immobile and the gas migrates to grain boundaries by diffusion of single gas atoms. The intragranular bubble population in the model at low burn-ups or temperatures consists of numerous small bubbles. The presence of the bubbles attenuates the effective gas atom diffusion coefficient. Rapid coarsening of the bubble population in increased burn-up at elevated temperatures weakens significantly the attenuation of the effective diffusion coefficient. The solution method introduced in earlier papers, locally accurate method, is enhanced to allow accurate calculation of the intragranular gas behaviour in time varying conditions without excessive computing time. Qualitatively the detailed model can predict the gas retention in the grain better than a more simple model

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

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Memmott, Matthew; Boy, Guy; Charit, Indrajit; Manera, Annalisa; Downar, Thomas; Lee, John; Muldrow, Lycurgus; Upadhyaya, Belle; Hines, Wesley; Haghighat, Alierza

    2017-01-01

    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 (I 2 S-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.

  14. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    International Nuclear Information System (INIS)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems

  15. Verification of the depletion capabilities of the MCNPX code on a LWR MOX fuel assembly

    International Nuclear Information System (INIS)

    Cerba, S.; Hrncir, M.; Necas, V.

    2012-01-01

    The study deals with the verification of the depletion capabilities of the MCNPX code, which is a linked Monte-Carlo depletion code. For such a purpose the IV-B phase of the OECD NEA Burnup credit benchmark has been chosen. The mentioned benchmark is a code to code comparison of the multiplication coefficient k eff and the isotopic composition of a LWR MOX fuel assembly at three given burnup levels and after five years of cooling. The benchmark consists of 6 cases, 2 different Pu vectors and 3 geometry models, however in this study only the fuel assembly calculations with two Pu vectors were performed. The aim of this study was to compare the obtained result with data from the participants of the OECD NEA Burnup Credit project and confirm the burnup capability of the MCNPX code. (Authors)

  16. Overview of LWR severe accident research activities at the Karlsruhe Institute of Technology

    International Nuclear Information System (INIS)

    Miassoedov, Alexei; Albrecht, Giancarlo; Foit, Jerzy-Jan; Jordan, Thomas; Steinbrück, Martin; Stuckert, Juri; Tromm, Walter

    2012-01-01

    The research activities in the light water reactor (LWR) severe accidents domain at Karlsruhe Institute of Technology (KIT) are concentrated on the in- and ex-vessel core melt behavior. The overall objective is to investigate the core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity and to the containment, corium concrete interaction and corium coolability in the reactor cavity, and hydrogen behaviour in reactor systems. The results of the experiments contribute to a better understanding of the core melt sequences and thus improve safety of existing and, in the long-term, of future reactors by severe accident mitigation measures and by safety installations where required. This overview paper describes the experimental facilities used at KIT for severe accident research and gives an overview of the main directions and objectives of the R&D work. (author)

  17. Assessment of management alternatives for LWR wastes. Volume 1. Main achievements of the joint study

    International Nuclear Information System (INIS)

    Glibert, R.C.

    1993-01-01

    This report deals with the main achievements of a joint theoretical study aimed at evaluating a selection of management routes for LWR wastes, relying to a certain extent on national practices in this particular area, on the basis of economical and radiological criteria. All individual intermediate steps entering a management route, from radioactive-wastes production up to their disposal in near-surface sites or in a deep repository, have been identified, described and cost-evaluated throughout the study. The radiological impact assessment comprises estimates of both individual and collective doses resulting from normal discharges of radioactive effluents and from disposal of radioactive waste products in near-surfaces sites. All specific data concerning the description of the different management routes considered as well as the methodology applied to evaluate cost and radiological impact are detailed in the subsequent volumes of the series (Volumes 2 to 8)

  18. Cracking in LWR RPV head penetrations. Working material. Proceedings of a specialists meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The IAEA Specialists` Meeting on Cracking in LWR RPV Head Penetrations was held at the ASTM Headquarters, Philadelphia, Pennsylvania, on May 2-4, 1995. It was attended by 39 participants from 12 countries. The meeting was held in the framework of the IAEA International Working Group on Life Management of Nuclear Power Plants (IWG-LMNPP) and was organized and sponsored by the Oak Ridge National Laboratory and the U.S. Nuclear Regulatory Commission. The purpose of the meeting was to review experience in the field for ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. Presentations were aimed at achieving a better understanding of the behaviour of reactor component materials, providing guidance and recommendations to assure reliability and adequate performance, and proposing directions for further investigations. Refs, figs and tabs.

  19. Critical corrosion issues and mitigation strategies impacting the operability of LWR's

    International Nuclear Information System (INIS)

    Jones, R.L.

    1996-01-01

    Recent corrosion experience in US light water reactor nuclear power plants is reviewed with emphasis on mitigation strategies to control the cost of corrosion to LWR operators. Many components have suffered corrosion problems resulting in industry costs of billions of dollars. The most costly issues have been stress corrosion cracking of stainless steel coolant piping in boiling water reactors and corrosion damage to steam generator tubes in pressurized water reactors. Through industry wide R and D programs these problems are now understood and mitigation strategies have been developed to address the issues in a cost effective manner. Other significant corrosion problems for both reactor types are briefly reviewed. Tremendous progress has been made in controlling corrosion, however, minimizing its impact on plant operations will present a continuing challenge throughout the remaining service lives of these power plants

  20. Evaluation of alternative waste management schemes for LWR hulls and caps

    International Nuclear Information System (INIS)

    Chaudon, L.; Cecille, L.; Klein, M.; Kowa, S.; Mehling, O.; Thiels, G.

    1990-01-01

    LWR hulls and caps represent one of the major sources of α bearing solid waste generated in the nuclear fuel cycle. For this reason, the CEC launched a theoretical study to evaluate alternative schemes for the overall management of this waste. Both volume reduction techniques and α decontamination of the hulls were assessed. The study demonstrated that the transport and disposal of the conditioned waste in deep geological formations play a dominant part in the total management costs. Important cost savings can be achieved through the implementation of efficient volume reduction techniques, i.e. melting or compaction. As an alternative approach, exhaustive α decontamination of the hulls appears promising, provided that the conditioned waste can be made to comply with the disposal criteria of mines. Finally, prolongation of the interim storage period for the waste packages from 1 to 30 years may prove beneficial on the transport costs

  1. 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.

  2. Improving the computation efficiency of COBRA-TF for LWR safety analysis of large problems

    International Nuclear Information System (INIS)

    Cuervo, D.; Avramova, M. N.; Ivanov, K. N.

    2004-01-01

    A matrix solver is implemented in COBRA-TF in order to improve the computation efficiency of both numerical solution methods existing in the code, the Gauss elimination and the Gauss-Seidel iterative technique. Both methods are used to solve the system of pressure linear equations and relay on the solution of large sparse matrices. The introduced solver accelerates the solution of these matrices in cases of large number of cells. The execution time is reduced in half as compared to the execution time without using matrix solver for the cases with large matrices. The achieved improvement and the planned future work in this direction are important for performing efficient LWR safety analyses of large problems. (authors)

  3. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.; Kabele, T.J.

    1979-09-01

    To contribute to the Department of Energy's identification of needs for improved environmental controls in nuclear fuel cycles, a study was made of a light water reactor system. A reference LWR fuel cycle was defined, and each step in this cycle was characterized by facility description and mainline and effluent treatment process performance. The reference fuel cycle uses fresh uranium in light water reactors. Final treatment and ultimate disposition of waste from the fuel cycle steps were not included, and the waste is assumed to be disposed of by approved but currently undefined means. The characterization of the reference fuel cycle system is intended as basic information for further evaluation of alternative effluent control systems.

  4. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    Energy Technology Data Exchange (ETDEWEB)

    Kristine Barrett; Shannon Bragg-Sitton

    2012-09-01

    The Advanced Light Water Reactor (LWR) Nuclear Fuel Development Research and Development (R&D) Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.

  5. Recent results from CEC cost sharing research programme on LWR fuel behaviour under accident conditions

    International Nuclear Information System (INIS)

    Fairbairn, S.A.

    1983-01-01

    The present structure and intentions of the CEC sponsored cost sharing programme for LWR safety research are outlined. Detailed results are reported for two projects from this programme. The first project concerns experimental data on the thermohydraulic effects of flow diversion around ballooned fuel rods. Data are presented on single and two phase heat transfer in an electrically heated rod bundle. Detailed photographic data on droplet behaviour are also given. The second project is an investigation of the effects of zircaloy oxidation on rewetting during reflood. It is shown that as oxide thickness increases from 1μm to 76μm that rewet rates can increase by up to 40%. A systematic effect of oxidation on rewet temperatures is also noted. (author)

  6. Cracking in LWR RPV head penetrations. Working material. Proceedings of a specialists meeting

    International Nuclear Information System (INIS)

    1995-01-01

    The IAEA Specialists' Meeting on Cracking in LWR RPV Head Penetrations was held at the ASTM Headquarters, Philadelphia, Pennsylvania, on May 2-4, 1995. It was attended by 39 participants from 12 countries. The meeting was held in the framework of the IAEA International Working Group on Life Management of Nuclear Power Plants (IWG-LMNPP) and was organized and sponsored by the Oak Ridge National Laboratory and the U.S. Nuclear Regulatory Commission. The purpose of the meeting was to review experience in the field for ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. Presentations were aimed at achieving a better understanding of the behaviour of reactor component materials, providing guidance and recommendations to assure reliability and adequate performance, and proposing directions for further investigations. Refs, figs and tabs

  7. Nonlinear analysis of LWR components: areas of investigation/benefits/recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, S. J. [ed.

    1980-04-01

    The purpose of this study is to identify specific topics of investigation into design procedures, design concepts, methods of analysis, testing practices, and standards which are characterized by nonlinear behavior (both geometric and material) and which are considered to offer some economic and/or technical benefits to the LWR industry (excluding piping). In this study these topics were collected, compiled, and subjectively evaluated as to their potential benefit. The topics considered to have the greatest benefit/impact potential are discussed. The topics of investigation were found to fall basically into three areas: component, code interpretation, and load/failure mechanism. The topics are arbitrarily reorganized into six areas of investigation: Fracture, Fatigue, Vibration/Dynamic/Seismic, Plasticity, Component/Computational Considerations, and Code Interpretation.

  8. Mechanical damage due to corrosion of parts of pump technology and valves of LWR power installations

    International Nuclear Information System (INIS)

    Hron, J.; Krumpl, M.

    1986-01-01

    Two types are described of uneven corrosion of austenitic chromium-nickel steel: pitting and slit corrosion. The occurrence of slit corrosion is typical of parts of pumping technology and valves. The corrosion damage of austenitic chromium-nickel steels spreads as intergranular, transgranular or mixed corrosion. In nuclear power facilities with LWR's, intergranular corrosion is due to chlorides and sulphur compounds while transgranular corrosion is due to the presence of dissolved oxygen and chlorides. In mechanically stressed parts, stress corrosion takes place. The recommended procedures are discussed of reducing the corrosion-mechanical damage of pumping equipment of light water reactors during design, production and assembly. During the service of the equipment, corrosion cracks are detected using nondestructive methods and surface cracks are repaired by grinding and welding. (E.S.)

  9. Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect

    International Nuclear Information System (INIS)

    Ishiguro, Misako; Harada, Hiroo; Shinozawa, Naohisa; Naraoka, Ken-itsu

    1985-03-01

    The RELAP5/MOD1 is a large thermal-hydraulic code to analyze LWR LOCA and non-LOCA transients. The code originally was designed for use on a CDC Cyber-176. This report documents vectorization of the RELAP5/MOD1 code conducted for the purpose of efficient use of VP-100 (peak speed 250 MFLOPS, clock period 7.5 ns) at the JAERI. The code was vectorized using the junction and volume level parallelisms in the hydrodynamic calculations, and the heat-structure and heat-mesh level in the heat conduction calculations. The vectorized version runs as much as 2.4 to 2.8 times faster than the original scalar version, while the speedup ratio is dependent on the number of spactial cells included in the problem. (author)

  10. Plant for retention of 14C in reprocessing plants for LWR fuel elements

    International Nuclear Information System (INIS)

    Braun, H.; Gutowski, H.; Bonka, H.; Gruendler, D.

    1983-01-01

    The 14 C produced from nuclear power plants is actually totally emitted from nuclear power plants and reprocessing plants. Using the radiation protection principles proposed in ICRP 26, 14 C should be retained at heavy water moderated reactors and reprocessing plants due to a cost-benefit analysis. In the frame of a research work to cost-benefit analysis, which was sponsored by the Federal Minister of the Interior, an industrial plant for 14 C retention at reprocessing plants for LWR fuel elements has been planned according to the double alkali process. The double alkali process has been chosen because of the sufficient operation experience in the conventional chemical technique. In order to verify some operational parameters and to gain experiences, a cold test plant was constructed. The experiment results showed that the double alkali process is a technically suitable method with high operation security. Solidifying CaCO 3 with cement gives a product fit for final disposal

  11. Aging assessment and mitigation for major LWR [light water reactor] components

    International Nuclear Information System (INIS)

    Shah, Y.N.; Ware, A.G.; Conley, D.A.; MacDonald, P.E.; Burns, J.J. Jr.

    1989-01-01

    This paper summarizes some of the results of the Aging Assessment and Mitigation Project sponsored by the US Nuclear Regulatory Commission (USNRC), Office of Nuclear Regulatory Research. The objective of the project is to develop an understanding of the aging degradation of the major light water reactor (LWR) structures and components and to develop methods for predicting the useful life of these components so that the impact of aging on the safe operation of nuclear power plants can be evaluated and addressed. The research effort consists of integrating, evaluating, and updating the available aging-related information. This paper discusses current accomplishments and summarizes the significant degradation processes active in two major components: pressurized water reactor pressurizer surge and spray lines and nozzles, and light water reactor primary coolant pumps. This paper also evaluates the effectiveness of the current inservice inspection programs and presents conclusions and recommendations related to aging of these two major components. 37 refs., 7 figs., 3 tabs

  12. Challenge Based Innovation: Translating Fundamental Research into Societal Applications

    Science.gov (United States)

    Kurikka, Joona; Utriainen, Tuuli; Repokari, Lauri

    2016-01-01

    This paper is based on work done at IdeaSquare, a new innovation experiment at CERN, the European Organization for Nuclear Research. The paper explores the translation of fundamental research into societal applications with the help of multidisciplinary student teams, project- and problem-based learning and design thinking methods. The theme is…

  13. Practical applications of the geographic coordinate data base in Arkansas

    Science.gov (United States)

    Mickie Warwick; Don C. Bragg

    2005-01-01

    Though not intended for these applications, the General Land Office (GLO) survey notes are a primary source of historical, ecological, and cultural information, making it imperative that their spatial coordinates be as reliable as possible. The Geographic Coordinate Data Base (GCDB) is a statistically-based coordinate fitting program that uses the GLO notes and other...

  14. Towards an MDA-based development methodology for distributed applications

    NARCIS (Netherlands)

    van Sinderen, Marten J.; Gavras, A.; Belaunde, M.; Ferreira Pires, Luis; Andrade Almeida, João

    2004-01-01

    This paper proposes a development methodology for distributed applications based on the principles and concepts of the Model-Driven Architecture (MDA). The paper identifies phases and activities of an MDA-based development trajectory, and defines the roles and products of each activity in accordance

  15. U.S. Forward Operating Base Applications of Nuclear Power

    International Nuclear Information System (INIS)

    Griffith, George W.

    2015-01-01

    This paper provides a high level overview of current nuclear power technology and the potential use of nuclear power at military bases. The size, power ranges, and applicability of nuclear power units for military base power are reviewed. Previous and current reactor projects are described to further define the potential for nuclear power for military power.

  16. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  17. Assessment of nitrogen as an atmosphere for dry storage of spent LWR fuel

    International Nuclear Information System (INIS)

    Gilbert, E.R.; Knox, C.A.; White, G.D.

    1985-09-01

    Interim dry storage of spent light-water reactor (LWR) fuel is being developed as a licensed technology in the United States. Because it is anticipated that license agreements will specify dry storage atmospheres, the behavior of spent LWR fuel in a nitrogen atmosphere during dry storage was investigated. In particular, the thermodynamics of reaction of nitrogen compounds (expected to form in the cover gas during dry storage) and residual impurities (such as moisture and oxygen) with Zircaloy cladding and with spent fuel at sites of cladding breaches were examined. The kinetics of reaction were not considered it was assumed that the 20 to 40 years of interim dry storage would be sufficient for reactions to proceed to completion. The primary thermodynamics reactants were found to be NO 2 , N 2 O, H 2 O 2 , and O 2 . The evaluation revealed that the limited inventories of these reactants produced by the source terms in hermetically sealed dry storage systems would be too low to cause significant spent fuel degradation. Furthermore, the oxidation of spent fuel to degrading O/U ratios is unlikely because the oxidation potential in moist nitrogen limits O/U ratios to values less than UO/sub 2.006/ (the equilibrium stoichiometric form in equilibrium with moist nitrogen). Tests were performed with bare spent UO 2 fuel and nonirradiated UO 2 pellets (with no Zircaloy cladding) in a nitrogen atmosphere containing moisture concentrations greater than encountered under dry storage conditions. These tests were performed for at least 1100 h at temperatures as high as 380 0 C, where oxidation reactions proceed in a matter of minutes. No visible degradation was detected, and weight changes were negligible

  18. New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

    International Nuclear Information System (INIS)

    Bomboni, Eleonora

    2008-01-01

    The long-term radiotoxicity of the final waste is currently the main drawback of nuclear power production. Particularly, isotopes of Neptunium and Plutonium along with some long-lived fission products are dangerous for more than 100000 years. 96% of spent Light Water Reactor (LWR) fuel consists of actinides, hence it is able to produce a lot of energy by fission if recycled. Goals of Generation IV Initiative are reduction of long-term radiotoxicity of waste to be stored in geological repositories, a better exploitation of nuclear fuel resources and proliferation resistance. Actually, all these issues are intrinsically connected with each other. It is quite clear that these goals can be achieved only by combining different concepts of Gen. IV nuclear cores in a 'symbiotic' way. Light-Water Reactor - (Very) High Temperature Reactor ((V)HTR) - Fast Reactor (FR) symbiotic cycles have good capabilities from the viewpoints mentioned above. Particularly, HTR fuelled by Plutonium oxide is able to reach an ultra-high burn-up and to burn Neptunium and Plutonium effectively. In contrast, not negligible amounts of Americium and Curium build up in this core, although the total mass of Heavy Metals (HM) is reduced. Americium and Curium are characterised by an high radiological hazard as well. Nevertheless, at least Plutonium from HTR (rich in non-fissile nuclides) and, if appropriate, Americium can be used as fuel for Fast Reactors. If necessary, dedicated assemblies for Minor Actinides (MA) burning can be inserted in Fast Reactors cores. This presentation focuses on combining HTR and Gas Cooled Fast Reactor (GCFR) concepts, fuelled by spent LWR fuel and depleted uranium if need be, to obtain a net reduction of total mass and radiotoxicity of final waste. The intrinsic proliferation resistance of this cycle is highlighted as well. Additionally, some hints about possible Curium management strategies are supplied. Besides, a preliminary assessment of different chemical forms of

  19. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  20. Development of a Web-based financial application System

    Science.gov (United States)

    Hasan, M. R.; Ibrahimy, M. I.; Motakabber, S. M. A.; Ferdaus, M. M.; Khan, M. N. H.; Mostafa, M. G.

    2013-12-01

    The paper describes a technique to develop a web based financial system, following latest technology and business needs. In the development of web based application, the user friendliness and technology both are very important. It is used ASP .NET MVC 4 platform and SQL 2008 server for development of web based financial system. It shows the technique for the entry system and report monitoring of the application is user friendly. This paper also highlights the critical situations of development, which will help to develop the quality product.

  1. DEVELOPMENT OF A WEB-BASED PROXIMITY BASED MEDIA SHARING APPLICATION

    OpenAIRE

    Erol Ozan

    2016-01-01

    This article reports the development of Vissou, which is a location based web application that enables media recording and sharing among users who are in close proximity to each other. The application facilitates the automated hand-over of the recorded media files from one user to another. There are many social networking applications and web sites that provide digital media sharing and editing functionalities. What differentiates Vissou from other similar applications is the functions and us...

  2. MEMS-Based Power Generation Techniques for Implantable Biosensing Applications

    Directory of Open Access Journals (Sweden)

    Jonathan Lueke

    2011-01-01

    Full Text Available Implantable biosensing is attractive for both medical monitoring and diagnostic applications. It is possible to monitor phenomena such as physical loads on joints or implants, vital signs, or osseointegration in vivo and in real time. Microelectromechanical (MEMS-based generation techniques can allow for the autonomous operation of implantable biosensors by generating electrical power to replace or supplement existing battery-based power systems. By supplementing existing battery-based power systems for implantable biosensors, the operational lifetime of the sensor is increased. In addition, the potential for a greater amount of available power allows additional components to be added to the biosensing module, such as computational and wireless and components, improving functionality and performance of the biosensor. Photovoltaic, thermovoltaic, micro fuel cell, electrostatic, electromagnetic, and piezoelectric based generation schemes are evaluated in this paper for applicability for implantable biosensing. MEMS-based generation techniques that harvest ambient energy, such as vibration, are much better suited for implantable biosensing applications than fuel-based approaches, producing up to milliwatts of electrical power. High power density MEMS-based approaches, such as piezoelectric and electromagnetic schemes, allow for supplemental and replacement power schemes for biosensing applications to improve device capabilities and performance. In addition, this may allow for the biosensor to be further miniaturized, reducing the need for relatively large batteries with respect to device size. This would cause the implanted biosensor to be less invasive, increasing the quality of care received by the patient.

  3. MEMS-based power generation techniques for implantable biosensing applications.

    Science.gov (United States)

    Lueke, Jonathan; Moussa, Walied A

    2011-01-01

    Implantable biosensing is attractive for both medical monitoring and diagnostic applications. It is possible to monitor phenomena such as physical loads on joints or implants, vital signs, or osseointegration in vivo and in real time. Microelectromechanical (MEMS)-based generation techniques can allow for the autonomous operation of implantable biosensors by generating electrical power to replace or supplement existing battery-based power systems. By supplementing existing battery-based power systems for implantable biosensors, the operational lifetime of the sensor is increased. In addition, the potential for a greater amount of available power allows additional components to be added to the biosensing module, such as computational and wireless and components, improving functionality and performance of the biosensor. Photovoltaic, thermovoltaic, micro fuel cell, electrostatic, electromagnetic, and piezoelectric based generation schemes are evaluated in this paper for applicability for implantable biosensing. MEMS-based generation techniques that harvest ambient energy, such as vibration, are much better suited for implantable biosensing applications than fuel-based approaches, producing up to milliwatts of electrical power. High power density MEMS-based approaches, such as piezoelectric and electromagnetic schemes, allow for supplemental and replacement power schemes for biosensing applications to improve device capabilities and performance. In addition, this may allow for the biosensor to be further miniaturized, reducing the need for relatively large batteries with respect to device size. This would cause the implanted biosensor to be less invasive, increasing the quality of care received by the patient.

  4. Application of an intermediate LWR for electricity production and hot-water district heating

    International Nuclear Information System (INIS)

    1983-05-01

    The objective of the study is to evaluate the technical and economic feasibility of a 400 MWe Consolidated Nuclear Steam System (CNSS) for supplying district heat to the Minneapolis/St. Paul area. A total of three CNSS reactor sites, located various distances from the Minneapolis-St. Paul area load center, are evaluated. The distance from the load center is determined by the credited safety features of the plant design. Each site is also evaluated for three different hot water supply/return temperatures providing a total of nine CNSS study cases. The cost of district heat delivered to the load center is determined for each case

  5. Development and application of PIE apparatuses for high-burnup LWR fuels

    International Nuclear Information System (INIS)

    Harada, Katsuya; Mita, Naoaki; Nishino, Yasuharu; Amano, Hidetoshi

    1999-01-01

    The Reactor Fuel Examination Facility (RFEF) is developing the following post irradiation examination apparatuses: Ion Microprobe mass analyzer (IMA), Pellet Thermal Capacity measuring apparatus (PTC), Micro Density Measuring apparatus MDM, Shield-type Field Emission Scanning Electron Microscope (FE-SEM). The present paper mainly describes several technical topics of these apparatuses. (author)

  6. Applications of the thermit code to 3D thermal hydraulic analysis of LWR cores

    International Nuclear Information System (INIS)

    Reed, W.H.

    1979-01-01

    The THERMIT code calculates the three-dimensional transient thermal hydraulic behavior of light water reactor cores. Its two-fluid dynamics equations for two-phase flow offer improved physical modelling capability needed in the context of calculation coupled to neutron kinetics for feedback. The numerical fluid dynamics method was chosen for reliability over a wider range of transients. An improved heat transfer numerical method is presented which gives better numerical stability and accuracy. A number of example calculations are discussed which give an idea of the power and flexibility of the code

  7. Collaborative Working Architecture for IoT-Based Applications.

    Science.gov (United States)

    Mora, Higinio; Signes-Pont, María Teresa; Gil, David; Johnsson, Magnus

    2018-05-23

    The new sensing applications need enhanced computing capabilities to handle the requirements of complex and huge data processing. The Internet of Things (IoT) concept brings processing and communication features to devices. In addition, the Cloud Computing paradigm provides resources and infrastructures for performing the computations and outsourcing the work from the IoT devices. This scenario opens new opportunities for designing advanced IoT-based applications, however, there is still much research to be done to properly gear all the systems for working together. This work proposes a collaborative model and an architecture to take advantage of the available computing resources. The resulting architecture involves a novel network design with different levels which combines sensing and processing capabilities based on the Mobile Cloud Computing (MCC) paradigm. An experiment is included to demonstrate that this approach can be used in diverse real applications. The results show the flexibility of the architecture to perform complex computational tasks of advanced applications.

  8. Emerging Applications of Liquid Crystals Based on Nanotechnology

    Directory of Open Access Journals (Sweden)

    Jung Inn Sohn

    2014-03-01

    Full Text Available Diverse functionalities of liquid crystals (LCs offer enormous opportunities for their potential use in advanced mobile and smart displays, as well as novel non-display applications. Here, we present snapshots of the research carried out on emerging applications of LCs ranging from electronics to holography and self-powered systems. In addition, we will show our recent results focused on the development of new LC applications, such as programmable transistors, a transparent and active-type two-dimensional optical array and self-powered display systems based on LCs, and will briefly discuss their novel concepts and basic operating principles. Our research will give insights not only into comprehensively understanding technical and scientific applications of LCs, but also developing new discoveries of other LC-based devices.

  9. THE DIFFERENCE BETWEEN DEVELOPING SINGLE PAGE APPLICATION AND TRADITIONAL WEB APPLICATION BASED ON MECHATRONICS ROBOT LABORATORY ONAFT APPLICATION

    Directory of Open Access Journals (Sweden)

    V. Solovei

    2018-04-01

    Full Text Available Today most of desktop and mobile applications have analogues in the form of web-based applications.  With evolution of development technologies and web technologies web application increased in functionality to desktop applications. The Web application consists of two parts of the client part and the server part. The client part is responsible for providing the user with visual information through the browser. The server part is responsible for processing and storing data.MPA appeared simultaneously with the Internet. Multiple-page applications work in a "traditional" way. Every change eg. display the data or submit data back to the server. With the advent of AJAX, MPA learned to load not the whole page, but only a part of it, which eventually led to the appearance of the SPA. SPA is the principle of development when only one page is transferred to the client part, and the content is downloaded only to a certain part of the page, without rebooting it, which allows to speed up the application and simplify the user experience of using the application to the level of desktop applications.Based on the SPA, the Mechatronics Robot Laboratory ONAFT application was designed to automate the management process. The application implements the client-server architecture. The server part consists of a RESTful API, which allows you to get unified access to the application functionality, and a database for storing information. Since the client part is a spa, this allows you to reduce the load on the connection to the server and improve the user experience

  10. QoS testing of service-based applications

    OpenAIRE

    Driss , Maha; Jamoussi , Yassine; Ben Ghézala , Henda Hajjami

    2008-01-01

    International audience; Web services (WSs) are becoming increasingly popular because of their potential in several application domains including e-Enterprise, e-Business, e-Government, and e-Science. Based on open XML standards, WS technology allows the construction of massively distributed and loosely coupled applications. A service composition mechanism should satisfy not only functional properties but also non-functional Quality of Service (QoS) ones. In this paper, we introduce a discrete...

  11. Development and Implementation of Location Based Native Mobile Application

    OpenAIRE

    Čović, Zlatko; Babić, Dejan

    2014-01-01

    This paper presents the development and implementation of a location based native mobile application for smart phones with the Android operating system. It can be used in promotion of some product or brand in the way of mobile games. The application has the administration interface (back-end) for data manipulation of users, administrators, clients and items. The system distinguishes between three levels of users: the main administrator, administrators-customers and users. The paper also descr...

  12. Development of Polymethylmethacrylate Based Composite for Gas Sensing Application

    OpenAIRE

    Devikala, S.; Kamaraj, P.

    2011-01-01

    Gas detection instruments are increasingly needed for industrial health and safety, environmental monitoring and process control. Conductive polymer composites have various industrial applications. The composite prepared by mixing carbon black with polymethylmethacrylate (PMMA) has very good gas sensing applications. The gas sensors based on carbon nanotube/polymer, ceramic and metal oxide composites such as epoxy, polyimide, PMMA / Barium titanate and tin oxide have also been developed. In t...

  13. Vanadium-base alloys for fusion reactor applications

    International Nuclear Information System (INIS)

    Smith, D.L.; Loomis, B.A.; Diercks, D.R.

    1984-10-01

    Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined

  14. Vanadium-base alloys for fusion reactor applications

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Loomis, B.A.; Diercks, D.R.

    1984-10-01

    Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined.

  15. Laser wakefield accelerator based light sources: potential applications and requirements

    Energy Technology Data Exchange (ETDEWEB)

    Albert, F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States). NIF and Photon Sciences; Thomas, A. G. [Univ. of Michigan, Ann Arbor, MI (United States). Dept. of Nuclear Engineering and Radiological Sciences; Mangles, S. P.D. [Imperial College, London (United Kingdom). Blackett Lab.; Banerjee, S. [Univ. of Nebraska, Lincoln, NE (United States); Corde, S. [SLAC National Accelerator Lab., Menlo Park, CA (United States); Flacco, A. [ENSTA, CNRS, Ecole Polytechnique, Palaiseau (France); Litos, M. [SLAC National Accelerator Lab., Menlo Park, CA (United States); Neely, D. [Science and Technology Facilities Council (STFC), Oxford (United Kingdom). Rutherford Appleton Lab. (RAL). Central Laser Facility; Viera, J. [Univ. of Lisbon (Portugal). GoLP-Inst. de Plasmas e Fusao Nuclear-Lab. Associado; Najmudin, Z. [Imperial College, London (United Kingdom). Blackett Lab.; Bingham, R. [Science and Technology Facilities Council (STFC), Oxford (United Kingdom). Rutherford Appleton Lab. (RAL). Central Laser Facility; Joshi, C. [Univ. of California, Los Angeles, CA (United States). Dept. of Electrical Engineering; Katsouleas, T. [Duke Univ., Durham, NC (United States). Platt School of Engineering

    2015-01-15

    In this article we review the prospects of laser wakefield accelerators as next generation light sources for applications. This work arose as a result of discussions held at the 2013 Laser Plasma Accelerators Workshop. X-ray phase contrast imaging, X-ray absorption spectroscopy, and nuclear resonance fluorescence are highlighted as potential applications for laser-plasma based light sources. We discuss ongoing and future efforts to improve the properties of radiation from plasma betatron emission and Compton scattering using laser wakefield accelerators for these specific applications.

  16. Chitosan-Based Polymer Blends: Current Status and applications

    International Nuclear Information System (INIS)

    Hefian, E.A.E.; Nasef, M.M.

    2014-01-01

    This paper reviews the latest developments in chitosan-based blends and their potential applications in various fields. Various blends together with other derivatives, such as composites and graft copolymers, have been developed to overcome chitosans disadvantages, including poor mechanical properties and to improve its functionality towards specific applications. The progress made in blending chitosan with synthetic and natural polymers is presented. The versatility and unique characteristics, such as hydrophilicity, film-forming ability, biodegradability, biocompatibility, antibacterial activity and non-toxicity of chitosan has contributed to the successful development of various blends for medical, pharmaceutical, agricultural and environmental applications. (author)

  17. Access Control of Web and Java Based Applications

    Science.gov (United States)

    Tso, Kam S.; Pajevski, Michael J.; Johnson, Bryan

    2011-01-01

    Cyber security has gained national and international attention as a result of near continuous headlines from financial institutions, retail stores, government offices and universities reporting compromised systems and stolen data. Concerns continue to rise as threats of service interruption, and spreading of viruses become ever more prevalent and serious. Controlling access to application layer resources is a critical component in a layered security solution that includes encryption, firewalls, virtual private networks, antivirus, and intrusion detection. In this paper we discuss the development of an application-level access control solution, based on an open-source access manager augmented with custom software components, to provide protection to both Web-based and Java-based client and server applications.

  18. XML-Based Visual Specification of Multidisciplinary Applications

    Science.gov (United States)

    Al-Theneyan, Ahmed; Jakatdar, Amol; Mehrotra, Piyush; Zubair, Mohammad

    2001-01-01

    The advancements in the Internet and Web technologies have fueled a growing interest in developing a web-based distributed computing environment. We have designed and developed Arcade, a web-based environment for designing, executing, monitoring, and controlling distributed heterogeneous applications, which is easy to use and access, portable, and provides support through all phases of the application development and execution. A major focus of the environment is the specification of heterogeneous, multidisciplinary applications. In this paper we focus on the visual and script-based specification interface of Arcade. The web/browser-based visual interface is designed to be intuitive to use and can also be used for visual monitoring during execution. The script specification is based on XML to: (1) make it portable across different frameworks, and (2) make the development of our tools easier by using the existing freely available XML parsers and editors. There is a one-to-one correspondence between the visual and script-based interfaces allowing users to go back and forth between the two. To support this we have developed translators that translate a script-based specification to a visual-based specification, and vice-versa. These translators are integrated with our tools and are transparent to users.

  19. Design and Implementation of Campus Application APP Based on Android

    Science.gov (United States)

    dongxu, Zhu; yabin, liu; xian lei, PI; weixiang, Zhou; meng, Huang

    2017-07-01

    In this paper, "Internet + campus" as the entrance of the Android technology based on the application of campus design and implementation of Application program. Based on GIS(Geographic Information System) spatial database, GIS spatial analysis technology, Java development technology and Android development technology, this system server adopts the Model View Controller architectue to realize the efficient use of campus information and provide real-time information of all kinds of learning and life for campus student at the same time. "Fingertips on the Institute of Disaster Prevention Science and Technology" release for the campus students of all grades of life, learning, entertainment provides a convenient.

  20. Application of condition based maintenance to nuclear power plants

    International Nuclear Information System (INIS)

    Sonoda, Yukio; Nakano, Tomohito; Shimizu, Shunichi; Iida, Jun; Atomura, Masakazu; Abe, Masahiro

    2002-01-01

    Device Karte management system which supports application of condition based maintenance to nuclear power plants has been developed. The purpose of this system is to support maintenance personnel in device inspection scheduling based on operating condition monitoring and maintenance histories. There are four functions: field database, degradation estimation, inspection time decision and maintenance planning. The authors have been applying this system to dozens of devices of Onagawa Nuclear Power Station Unit No. 1 for one year. This paper represents the system concept and its application experiences. (author)