WorldWideScience

Sample records for line-width roughness lwr

  1. Effects of vacuum ultraviolet photons, ion energy and substrate temperature on line width roughness and RMS surface roughness of patterned 193 nm photoresist

    Titus, M J; Graves, D B; Yamaguchi, Y; Hudson, E A

    2011-01-01

    We present a comparison of patterned 193 nm photoresist (PR) line width roughness (LWR) of samples processed in a well characterized argon (Ar) inductively coupled plasma (ICP) system to RMS surface roughness and bulk chemical modification of blanket 193 nm PR samples used as control samples. In the ICP system, patterned and blanket PR samples are irradiated with Ar vacuum ultraviolet photons (VUV) and Ar ions while sample temperature, photon flux, ion flux and ion energy are controlled and measured. The resulting chemical modifications to bulk 193 nm PR (blanket) and surface roughness are analysed with Fourier transform infrared spectroscopy and atomic force microscopy (AFM). LWR of patterned samples are measured with scanning electron microscopy and blanket portions of the patterned PRs are measured with AFM. We demonstrate that with no RF-bias applied to the substrate the LWR of 193 nm PR tends to smooth and correlates with the smoothing of the RMS surface roughness. However, both LWR and RMS surface roughness increases with simultaneous high-energy (≥70 eV) ion bombardment and VUV-irradiation and is a function of exposure time. Both high- and low-frequency LWR correlate well with the RMS surface roughness of the patterned and blanket 193 nm PR samples. LWR, however, does not increase with temperatures ranging from 20 to 80 deg. C, in contrast to the RMS surface roughness which increases monotonically with temperature. It is unclear why LWR remains independent of temperature over this range. However, the fact that blanket roughness and LWR on patterned samples, both scale similarly with VUV fluence and ion energy suggests a similar mechanism is responsible for both types of surface morphology modifications.

  2. Effects of vacuum ultraviolet photons, ion energy and substrate temperature on line width roughness and RMS surface roughness of patterned 193 nm photoresist

    Titus, M J; Graves, D B [Department of Chemical Engineering, University of California, Berkeley, CA 94720 (United States); Yamaguchi, Y; Hudson, E A, E-mail: graves@berkeley.edu [Lam Research Corporation, 4400 Cushing Parkway, Freemont, CA 94538 (United States)

    2011-03-02

    We present a comparison of patterned 193 nm photoresist (PR) line width roughness (LWR) of samples processed in a well characterized argon (Ar) inductively coupled plasma (ICP) system to RMS surface roughness and bulk chemical modification of blanket 193 nm PR samples used as control samples. In the ICP system, patterned and blanket PR samples are irradiated with Ar vacuum ultraviolet photons (VUV) and Ar ions while sample temperature, photon flux, ion flux and ion energy are controlled and measured. The resulting chemical modifications to bulk 193 nm PR (blanket) and surface roughness are analysed with Fourier transform infrared spectroscopy and atomic force microscopy (AFM). LWR of patterned samples are measured with scanning electron microscopy and blanket portions of the patterned PRs are measured with AFM. We demonstrate that with no RF-bias applied to the substrate the LWR of 193 nm PR tends to smooth and correlates with the smoothing of the RMS surface roughness. However, both LWR and RMS surface roughness increases with simultaneous high-energy ({>=}70 eV) ion bombardment and VUV-irradiation and is a function of exposure time. Both high- and low-frequency LWR correlate well with the RMS surface roughness of the patterned and blanket 193 nm PR samples. LWR, however, does not increase with temperatures ranging from 20 to 80 deg. C, in contrast to the RMS surface roughness which increases monotonically with temperature. It is unclear why LWR remains independent of temperature over this range. However, the fact that blanket roughness and LWR on patterned samples, both scale similarly with VUV fluence and ion energy suggests a similar mechanism is responsible for both types of surface morphology modifications.

  3. An OCD perspective of line edge and line width roughness metrology

    Bonam, Ravi; Muthinti, Raja; Breton, Mary; Liu, Chi-Chun; Sieg, Stuart; Seshadri, Indira; Saulnier, Nicole; Shearer, Jeffrey; Patlolla, Raghuveer; Huang, Huai

    2017-03-01

    Metrology of nanoscale patterns poses multiple challenges that range from measurement noise, metrology errors, probe size etc. Optical Metrology has gained a lot of significance in the semiconductor industry due to its fast turn around and reliable accuracy, particularly to monitor in-line process variations. Apart from monitoring critical dimension, thickness of films, there are multiple parameters that can be extracted from Optical Metrology models3. Sidewall angles, material compositions etc., can also be modeled to acceptable accuracy. Line edge and Line Width roughness are much sought of metrology following critical dimension and its uniformity, although there has not been much development in them with optical metrology. Scanning Electron Microscopy is still used as a standard metrology technique for assessment of Line Edge and Line Width roughness. In this work we present an assessment of Optical Metrology and its ability to model roughness from a set of structures with intentional jogs to simulate both Line edge and Line width roughness at multiple amplitudes and frequencies. We also present multiple models to represent roughness and extract relevant parameters from Optical metrology. Another critical aspect of optical metrology setup is correlation of measurement to a complementary technique to calibrate models. In this work, we also present comparison of roughness parameters extracted and measured with variation of image processing conditions on a commercially available CD-SEM tool.

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

    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

  5. Stokes line width

    Nikiskov, A.I.; Ritus, V.I.

    1993-01-01

    The concept of Stokes line width is introduced for the asymptotic expansions of functions near an essential singularity. Explicit expressions are found for functions (switching functions) that switch on the exponentially small terms for the Dawson integral, Airy function, and the gamma function. A different, more natural representation of a function, not associated with expansion in an asymptotic series, in the form of dominant and recessive terms is obtained by a special division of the contour integral which represents the function into contributions of higher and lower saddle points. This division leads to a narrower, natural Stokes line width and a switching function of an argument that depends on the topology of the lines of steepest descent from the saddle point

  6. Line width of Josephson flux flow oscillators

    Koshelets, V.P.; Dmitriev, P.N.; Sobolev, A.S.

    2002-01-01

    to be proven before one initiates real FFO applications. To achieve this goal a comprehensive set of line width measurements of the FFO operating in different regimes has been performed. FFOs with tapered shape have been successfully implemented in order to avoid the superfine resonant structure with voltage...... spacing of about 20 nV and extremely low differential resistance, recently observed in the IVC of the standard rectangular geometry. The obtained results have been compared with existing theories and FFO models in order to understand and possibly eliminate excess noise in the FFO. The intrinsic line width...

  7. Experimental methodology of contact edge roughness on sub-100-nm pattern

    Lee, Tae Yong; Ihm, Dongchul; Kang, Hyo Chun; Lee, Jun Bum; Lee, Byoung-Ho; Chin, Soo-Bok; Cho, Do-Hyun; Kim, Yang Hyong; Yang, Ho Dong; Yang, Kyoung Mo

    2004-05-01

    The measurement of edge roughness has become a hot issue in the semiconductor industry. Major vendors offer a variety of features to measure the edge roughness in their CD-SEMs. However, most of the features are limited by the applicable pattern types. For the line and space patterns, features such as Line Edge Roughness (LER) and Line Width Roughness (LWR) are available in current CD-SEMs. The edge roughness is more critical in contact process. However the measurement of contact edge roughness (CER) or contact space roughness (CSR) is more complicated than that of LER or LWR. So far, no formal standard measurement algorithm or definition of contact roughness measurement exists. In this article, currently available features are investigated to assess their representability for CER or CSR. Some new ideas to quantify CER and CSR were also suggested with preliminary experimental results.

  8. Line Width Recovery after Vectorization of Engineering Drawings

    Gramblička Matúš

    2016-12-01

    Full Text Available Vectorization is the conversion process of a raster image representation into a vector representation. The contemporary commercial vectorization software applications do not provide sufficiently high quality outputs for such images as do mechanical engineering drawings. Line width preservation is one of the problems. There are applications which need to know the line width after vectorization because this line attribute carries the important semantic information for the next 3D model generation. This article describes the algorithm that is able to recover line width of individual lines in the vectorized engineering drawings. Two approaches are proposed, one examines the line width at three points, whereas the second uses a variable number of points depending on the line length. The algorithm is tested on real mechanical engineering drawings.

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

    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.

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

    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.

  11. Fingerprinting the type of line edge roughness

    Fernández Herrero, A.; Pflüger, M.; Scholze, F.; Soltwisch, V.

    2017-06-01

    Lamellar gratings are widely used diffractive optical elements and are prototypes of structural elements in integrated electronic circuits. EUV scatterometry is very sensitive to structure details and imperfections, which makes it suitable for the characterization of nanostructured surfaces. As compared to X-ray methods, EUV scattering allows for steeper angles of incidence, which is highly preferable for the investigation of small measurement fields on semiconductor wafers. For the control of the lithographic manufacturing process, a rapid in-line characterization of nanostructures is indispensable. Numerous studies on the determination of regular geometry parameters of lamellar gratings from optical and Extreme Ultraviolet (EUV) scattering also investigated the impact of roughness on the respective results. The challenge is to appropriately model the influence of structure roughness on the diffraction intensities used for the reconstruction of the surface profile. The impact of roughness was already studied analytically but for gratings with a periodic pseudoroughness, because of practical restrictions of the computational domain. Our investigation aims at a better understanding of the scattering caused by line roughness. We designed a set of nine lamellar Si-gratings to be studied by EUV scatterometry. It includes one reference grating with no artificial roughness added, four gratings with a periodic roughness distribution, two with a prevailing line edge roughness (LER) and another two with line width roughness (LWR), and four gratings with a stochastic roughness distribution (two with LER and two with LWR). We show that the type of line roughness has a strong impact on the diffuse scatter angular distribution. Our experimental results are not described well by the present modelling approach based on small, periodically repeated domains.

  12. Unbiased roughness measurements: the key to better etch performance

    Liang, Andrew; Mack, Chris; Sirard, Stephen; Liang, Chen-wei; Yang, Liu; Jiang, Justin; Shamma, Nader; Wise, Rich; Yu, Jengyi; Hymes, Diane

    2018-03-01

    Edge placement error (EPE) has become an increasingly critical metric to enable Moore's Law scaling. Stochastic variations, as characterized for lines by line width roughness (LWR) and line edge roughness (LER), are dominant factors in EPE and known to increase with the introduction of EUV lithography. However, despite recommendations from ITRS, NIST, and SEMI standards, the industry has not agreed upon a methodology to quantify these properties. Thus, differing methodologies applied to the same image often result in different roughness measurements and conclusions. To standardize LWR and LER measurements, Fractilia has developed an unbiased measurement that uses a raw unfiltered line scan to subtract out image noise and distortions. By using Fractilia's inverse linescan model (FILM) to guide development, we will highlight the key influences of roughness metrology on plasma-based resist smoothing processes. Test wafers were deposited to represent a 5 nm node EUV logic stack. The patterning stack consists of a core Si target layer with spin-on carbon (SOC) as the hardmask and spin-on glass (SOG) as the cap. Next, these wafers were exposed through an ASML NXE 3350B EUV scanner with an advanced chemically amplified resist (CAR). Afterwards, these wafers were etched through a variety of plasma-based resist smoothing techniques using a Lam Kiyo conductor etch system. Dense line and space patterns on the etched samples were imaged through advanced Hitachi CDSEMs and the LER and LWR were measured through both Fractilia and an industry standard roughness measurement software. By employing Fractilia to guide plasma-based etch development, we demonstrate that Fractilia produces accurate roughness measurements on resist in contrast to an industry standard measurement software. These results highlight the importance of subtracting out SEM image noise to obtain quicker developmental cycle times and lower target layer roughness.

  13. Real-time line-width measurements: a new feature for reticle inspection systems

    Eran, Yair; Greenberg, Gad; Joseph, Amnon; Lustig, Cornel; Mizrahi, Eyal

    1997-07-01

    The significance of line width control in mask production has become greater with the lessening of defect size. There are two conventional methods used for controlling line widths dimensions which employed in the manufacturing of masks for sub micron devices. These two methods are the critical dimensions (CD) measurement and the detection of edge defects. Achieving reliable and accurate control of line width errors is one of the most challenging tasks in mask production. Neither of the two methods cited above (namely CD measurement and the detection of edge defects) guarantees the detection of line width errors with good sensitivity over the whole mask area. This stems from the fact that CD measurement provides only statistical data on the mask features whereas applying edge defect detection method checks defects on each edge by itself, and does not supply information on the combined result of error detection on two adjacent edges. For example, a combination of a small edge defect together with a CD non- uniformity which are both within the allowed tolerance, may yield a significant line width error, which will not be detected using the conventional methods (see figure 1). A new approach for the detection of line width errors which overcomes this difficulty is presented. Based on this approach, a new sensitive line width error detector was developed and added to Orbot's RT-8000 die-to-database reticle inspection system. This innovative detector operates continuously during the mask inspection process and scans (inspects) the entire area of the reticle for line width errors. The detection is based on a comparison of measured line width that are taken on both the design database and the scanned image of the reticle. In section 2, the motivation for developing this new detector is presented. The section covers an analysis of various defect types, which are difficult to detect using conventional edge detection methods or, alternatively, CD measurements. In section 3

  14. Impacts of yttrium substitution on FMR line-width and magnetic properties of nickel spinel ferrites

    Ishaque, M., E-mail: ishaqdgk1@gmail.com [Department of Physics, Bahauddin Zakariya University, Multan 60800 (Pakistan); Khan, Muhammad Azhar, E-mail: azhar.khan@iub.edu.pk [Department of Physics, The Islamia University of Bahawalpur, Bahawalpur 63100 (Pakistan); Ali, Irshad; Khan, Hasan M. [Department of Physics, Bahauddin Zakariya University, Multan 60800 (Pakistan); Iqbal, M. Asif [Department of Physics, Bahauddin Zakariya University, Multan 60800 (Pakistan); College of E & ME, National University of Science and Technology, Islamabad (Pakistan); Islam, M.U. [Department of Physics, Bahauddin Zakariya University, Multan 60800 (Pakistan); Warsi, Muhammad Farooq [Department of Chemistry, The Islamia University of Bahawalpur, Bahawalpur 63100 (Pakistan)

    2015-05-15

    The influence of yttrium (Y) substitution on ferromagnetic resonance (FMR), initial permeability, and magnetic properties of NiFe{sub 2}O{sub 4} ferrites were investigated. It was observed that the FMR line-width decreases with yttrium contents for the substitution level 0≤×≤0.06. Beyond this, the FMR line-width increases with yttrium contents. The nominal composition NiY{sub 0.12}Fe{sub 1.88}O{sub 4} exhibited the smallest FMR line-width ~282 Oe. A significant change in FMR position of nickel–yttrium (Ni–Y) ferrites was observed and it found to exist between 4150 and 4600 Oe. The saturation magnetization was observed to decrease with the increase of yttrium contents and this was referred to the redistribution of cations on octahedral. The coercivity increased from 15 Oe to 59 Oe by increasing the yttrium concentration. The initial permeability decreased from 110 to 35 at 1 MHz by the incorporation of yttrium and this was attributed to the smaller grains which may obstruct the domain wall movement and impede the domain wall motion. The magnetic loss factors of substituted samples exhibit decreasing behavior in the frequency range 1 kHz to 10 MHz. The smaller FMR line-width and reduced magnetic loss factor of the investigated samples suggest the possible use of these materials in high frequency applications. - Highlights: • Influence of Y{sup 3+} substitution on the properties of nickel ferrites is investigated. • Very small FMR line-width (282 Oe) is exhibited by these substituted ferrites. • Fourfold increase in coercivity was observed for NiY{sub 0.24}Fe{sub 1.76}O{sub 4} ferrites.

  15. Experimental study of contact edge roughness on sub-100 nm various circular shapes

    Lee, Tae Y.; Ihm, Dongchul; Kang, Hyo C.; Lee, Jum B.; Lee, Byoung H.; Chin, Soo B.; Cho, Do H.; Song, Chang L.

    2005-05-01

    The measurement of edge roughness has become a hot issue in the semiconductor industry. Especially the contact roughness is being more critical as design rule shrinks. Major vendors offer a variety of features to measure the edge roughness in their CD-SEMs. For the line and space patterns, features such as Line Edge Roughness (LER) and Line Width Roughness (LWR) are available in current CD-SEMs. However the features currently available in commercial CD-SEM cannot provide a proper solution in monitoring the contact roughness. We had introduced a new parameter R, measurement algorithm and definition of contact edge roughness to quantify CER and CSR in previous paper. The parameter, R could provide an alternative solution to monitor contact or island pattern roughness. In this paper, we investigated to assess optimum number of CD measurement (1-D) and fitting method for CER or CSR. The study was based on a circular contact shape. Some new ideas to quantify CER or CSR were also suggested with preliminary experimental results.

  16. Line width measurement below 60 nm using an optical interferometer and artificial neural network

    See, Chung W.; Smith, Richard J.; Somekh, Michael G.; Yacoot, Andrew

    2007-03-01

    We have recently described a technique for optical line-width measurements. The system currently is capable of measuring line-width down to 60 nm with a precision of 2 nm, and potentially should be able to measure down to 10nm. The system consists of an ultra-stable interferometer and artificial neural networks (ANNs). The former is used to generate optical profiles which are input to the ANNs. The outputs of the ANNs are the desired sample parameters. Different types of samples have been tested with equally impressive results. In this paper we will discuss the factors that are essential to extend the application of the technique. Two of the factors are signal conditioning and sample classification. Methods, including principal component analysis, that are capable of performing these tasks will be considered.

  17. Emission-line widths and stellar-wind flows in T Tauri stars

    Sa, C.; Lago, M.T.V.T.

    1986-01-01

    Spectra are reported of T Tauri stars taken with the IPCS on the Isaac Newton Telescope at the Observatorio del Roque de los Muchachos at a dispersion of l7 A mm -1 . These were taken in order to determine emission-line widths and hence flow velocities in the winds of these stars following the successful modelling of the wind from RU Lupi using such data. Line widths in RW Aur suggest a similar pattern to the wind flow as in RU Lupi with velocities rising in the inner chromosphere of the star and then entering a 'ballistic' zone. The wind from DFTau is also similar but velocities are generally much lower and the lines sharper. (author)

  18. Maskless Lithography Using Negative Photoresist Material: Impact of UV Laser Intensity on the Cured Line Width

    Mohammed, Mohammed Ziauddin; Mourad, Abdel-Hamid I.; Khashan, Saud A.

    2018-06-01

    The application of maskless lithography technique on negative photoresist material is investigated in this study. The equipment used in this work is designed and built especially for maskless lithography applications. The UV laser of 405 nm wavelength with 0.85 Numerical Aperture is selected for direct laser writing. All the samples are prepared on a glass substrate. Samples are tested at different UV laser intensities and different stage velocities in order to study the impact on patterned line width. Three cases of spin coated layers of thickness 90 μm, 40 μm, and 28 μm on the substrate are studied. The experimental results show that line width has a generally increasing trend with intensity. However, a decreasing trend was observed for increasing velocity. The overall performance shows that the mr-DWL material is suitable for direct laser writing systems.

  19. Maskless Lithography Using Negative Photoresist Material: Impact of UV Laser Intensity on the Cured Line Width

    Mohammed, Mohammed Ziauddin; Mourad, Abdel-Hamid I.; Khashan, Saud A.

    2018-04-01

    The application of maskless lithography technique on negative photoresist material is investigated in this study. The equipment used in this work is designed and built especially for maskless lithography applications. The UV laser of 405 nm wavelength with 0.85 Numerical Aperture is selected for direct laser writing. All the samples are prepared on a glass substrate. Samples are tested at different UV laser intensities and different stage velocities in order to study the impact on patterned line width. Three cases of spin coated layers of thickness 90 μm, 40 μm, and 28 μm on the substrate are studied. The experimental results show that line width has a generally increasing trend with intensity. However, a decreasing trend was observed for increasing velocity. The overall performance shows that the mr-DWL material is suitable for direct laser writing systems.

  20. The effect of the negative binomial distribution on the line-width of the micromaser cavity field

    Kremid, A. M.

    2009-01-01

    The influence of negative binomial distribution (NBD) on the line-width of the negative binomial distribution (NBD) on the line-width of the micromaser is considered. The threshold of the micromaser is shifted towards higher values of the pumping parameter q. Moreover the line-width exhibits sharp dips 'resonances' when the cavity temperature reduces to a very low value. These dips are very clear evidence for the occurrence of the so-called trapping states regime in the micromaser. This statistics prevents the appearance of these trapping states, namely by increasing the negative binomial parameter q these dips wash out and the line-width becomes more broadening. For small values of the parameter q the line-width at large values of q randomly oscillates around its transition line. As q becomes large this oscillatory behavior occurs at rarely values of q. (author)

  1. Relating Line Width and Optical Depth for CO Emission in the Large Mgellanic Cloud

    Wojciechowski, Evan; Wong, Tony; Bandurski, Jeffrey; MC3 (Mapping CO in Molecular Clouds in the Magellanic Clouds) Team

    2018-01-01

    We investigate data produced from ALMA observations of giant molecular clouds (GMCs) located in the Large Magellanic Cloud (LMC), using 12CO(2–1) and 13CO(2–1) emission. The spectral line width is generally interpreted as tracing turbulent rather than thermal motions in the cloud, but could also be affected by optical depth, especially for the 12CO line (Hacar et al. 2016). We compare the spectral line widths of both lines with their optical depths, estimated from an LTE analysis, to evaluate the importance of optical depth effects. Our cloud sample includes two regions recently published by Wong et al. (2017, submitted): the Tarantula Nebula or 30 Dor, an HII region rife with turbulence, and the Planck cold cloud (PCC), located in a much calmer environment near the fringes of the LMC. We also include four additional LMC clouds, which span intermediate levels of star formation relative to these two clouds, and for which we have recently obtained ALMA data in Cycle 4.

  2. On the analysis of the thermal line shift and thermal line width of ions in solids

    Walsh, Brian M., E-mail: brian.m.walsh@nasa.gov [NASA Langley Research Center, Hampton, VA 23681 (United States); Di Bartolo, Baldassare, E-mail: baldassare.dibartolo@bc.edu [Boston College, Department of Physics, Chestnut Hill, MA 23667 (United States)

    2015-02-15

    A method of analysis for the thermally induced line shift and line width of spectral lines regarding the Raman process of ions in solids utilizing rational approximations for the Debye functions is presented. The {sup 2}E level unsplit R-line in V{sup 2+}:MgO is used as an example to illustrate the utility of the methods discussed here in providing a new analytical tool for researchers. - Highlights: • We use rational approximations for Debye functions. • We discuss limits and ranges of applicability of the rational approximations. • We formulate expressions for thermal shift and thermal linewidth for Raman processes using the rational approximations of the Debye functions. • We present an application of the methods to analyze the temperature dependent linewidth and lineshift in V2+:MgO.

  3. Determination of core level line widths in XPS of GeS and GeSe

    Viljoen, P.E.

    1981-01-01

    Measured X-ray photoelectron spectra are broadened owing to several factors. They can be regarded as the sums of the instrument response functions and the finite source widths. By measuring the response function and deconvoluting the measured peak, the form of the measured peak, the instrument function and the deconvoluted line were determined. The former two seem to have a Gauss and the latter a Lorentz form. The X-ray source is known to have a Lorentz form. A simple method, using the shapes of all these lines, is proposed to determine the line width. Applied to GeS and GeSe lines it gives values that agree quite well with other determinations. Strictly speaking, the method is only applicable to our or other similar spectrometers, but it can be generally applied if the line shapes are known or can be determined [af

  4. Roughness and uniformity improvements on self-aligned quadruple patterning technique for 10nm node and beyond by wafer stress engineering

    Liu, Eric; Ko, Akiteru; O'Meara, David; Mohanty, Nihar; Franke, Elliott; Pillai, Karthik; Biolsi, Peter

    2017-05-01

    Dimension shrinkage has been a major driving force in the development of integrated circuit processing over a number of decades. The Self-Aligned Quadruple Patterning (SAQP) technique is widely adapted for sub-10nm node in order to achieve the desired feature dimensions. This technique provides theoretical feasibility of multiple pitch-halving from 193nm immersion lithography by using various pattern transferring steps. The major concept of this approach is to a create spacer defined self-aligned pattern by using single lithography print. By repeating the process steps, double, quadruple, or octuple are possible to be achieved theoretically. In these small architectures, line roughness control becomes extremely important since it may contribute to a significant portion of process and device performance variations. In addition, the complexity of SAQP in terms of processing flow makes the roughness improvement indirective and ineffective. It is necessary to discover a new approach in order to improve the roughness in the current SAQP technique. In this presentation, we demonstrate a novel method to improve line roughness performances on 30nm pitch SAQP flow. We discover that the line roughness performance is strongly related to stress management. By selecting different stress level of film to be deposited onto the substrate, we can manipulate the roughness performance in line and space patterns. In addition, the impact of curvature change by applied film stress to SAQP line roughness performance is also studied. No significant correlation is found between wafer curvature and line roughness performance. We will discuss in details the step-by-step physical performances for each processing step in terms of critical dimension (CD)/ critical dimension uniformity (CDU)/line width roughness (LWR)/line edge roughness (LER). Finally, we summarize the process needed to reach the full wafer performance targets of LWR/LER in 1.07nm/1.13nm on 30nm pitch line and space pattern.

  5. The LER/LWR metrology challenge for advance process control through 3D-AFM and CD-SEM

    Faurie, P.; Foucher, J.; Foucher, A.-L.

    2009-12-01

    The continuous shrinkage in dimensions of microelectronic devices has reached such level, with typical gate length in advance R&D of less than 20nm combine with the introduction of new architecture (FinFET, Double gate...) and new materials (porous interconnect material, 193 immersion resist, metal gate material, high k materials...), that new process parameters have to be well understood and well monitored to guarantee sufficient production yield in a near future. Among these parameters, there are the critical dimensions (CD) associated to the sidewall angle (SWA) values, the line edge roughness (LER) and the line width roughness (LWR). Thus, a new metrology challenge has appeared recently and consists in measuring "accurately" the fabricated patterns on wafers in addition to measure the patterns on a repeatable way. Therefore, a great effort has to be done on existing techniques like CD-SEM, Scatterometry and 3D-AFM in order to develop them following the two previous criteria: Repeatability and Accuracy. In this paper, we will compare the 3D-AFM and CD-SEM techniques as a mean to measure LER and LWR on silicon and 193 resist and point out CD-SEM impact on the material during measurement. Indeed, depending on the material type, the interaction between the electron beam and the material or between the AFM tip and the material can vary a lot and subsequently can generate measurements bias. The first results tend to show that depending on CD-SEM conditions (magnification, number of acquisition frames) the final outputs can vary on a large range and therefore show that accuracy in such measurements are really not obvious to obtain. On the basis of results obtained on various materials that present standard sidewall roughness, we will show the limit of each technique and will propose different ways to improve them in order to fulfil advance roadmap requirements for the development of the next IC generation.

  6. NARROW-LINE-WIDTH UV BURSTS IN THE TRANSITION REGION ABOVE SUNSPOTS OBSERVED BY IRIS

    Hou, Zhenyong; Huang, Zhenghua; Xia, Lidong; Li, Bo; Madjarska, Maria S.; Fu, Hui; Mou, Chaozhou; Xie, Haixia, E-mail: z.huang@sdu.edu.cn, E-mail: xld@sdu.edu.cn [Shandong Provincial Key Laboratory of Optical Astronomy and Solar-Terrestrial Environment, Institute of Space Sciences, Shandong University, Weihai, 264209 Shandong (China)

    2016-10-01

    Various small-scale structures abound in the solar atmosphere above active regions, playing an important role in the dynamics and evolution therein. We report on a new class of small-scale transition region structures in active regions, characterized by strong emissions but extremely narrow Si iv line profiles as found in observations taken with the Interface Region Imaging Spectrograph (IRIS). Tentatively named as narrow-line-width UV bursts (NUBs), these structures are located above sunspots and comprise one or multiple compact bright cores at sub-arcsecond scales. We found six NUBs in two data sets (a raster and a sit-and-stare data set). Among these, four events are short-lived with a duration of ∼10 minutes, while two last for more than 36 minutes. All NUBs have Doppler shifts of 15–18 km s{sup −1}, while the NUB found in sit-and-stare data possesses an additional component at ∼50 km s{sup −1} found only in the C ii and Mg ii lines. Given that these events are found to play a role in the local dynamics, it is important to further investigate the physical mechanisms that generate these phenomena and their role in the mass transport in sunspots.

  7. Neonatal line width in deciduous incisors from Neolithic, mediaeval and modern skeletal samples from north-central Poland.

    Kurek, Marta; Żądzińska, Elżbieta; Sitek, Aneta; Borowska-Strugińska, Beata; Rosset, Iwona; Lorkiewicz, Wiesław

    2016-01-01

    The neonatal line is usually the first accentuated incremental line visible on the enamel. The prenatal environment significantly contributes to the width of the neonatal line, influencing the pace of reaching post-delivery homeostasis by the newborn's organism. Studies of the enamel of the earliest developing deciduous teeth can provide an insight into the prenatal development and the perinatal conditions of children of past human populations, thus being an additional source contributing to consideration of the influence of prenatal and perinatal factors modifying growth processes. The aim of this study was to examine whether the neonatal line, reflecting the conditions of the prenatal and perinatal environment, differed between the Neolithic, the mediaeval and the modern populations from the Kujawy region in north-central Poland. The material consisted of longitudinally ground sections of 57 human deciduous incisors obtained from children aged 1.0-7.5 years representing three archaeological series from Brześć Kujawski site. All teeth were sectioned in the labio-linqual plane using a diamond blade (Buechler IsoMet 1000). Final specimens were observed with the microscope Delta Optical Evolution 300 at 10× and 40× magnifications. For each tooth, linear measurements of the neonatal line width were performed on its labial surface at the three levels from the cemento-enamel junction. No significant difference was found in the mean neonatal line width depending on the tooth type and archaeological site, although the thickest neonatal line characterised children from the Neolithic series. In all analysed series, the neonatal line width was diversified depending on the child's age at death. The value of Spearman's rank correlation coefficient calculated for the correlation between the child's age at death and the neonatal line width was statistically significant. A clear increase in the width of the neonatal line was thus observed along with a decrease in the child

  8. Line width and line shape analysis in the inductively coupled plasma by high resolution Fourier transform spectrometry

    Faires, L.M.; Palmer, B.A.; Brault, J.W.

    1984-01-01

    High resolution Fourier transform spectrometry has been used to perform line width and line shape analysis of eighty-one iron I emision lines in the spectral range 290 to 390nm originating in the normal analytical zone of an inductively coupled plasma. Computer programs using non-linear least squares fitting techniques for line shape analysis were applied to the fully resolved spectra to determine Gaussian and Lorentzian components of the total observed line width. The effect of noise in the spectrum on the precision of the line fitting technique was assessed, and the importance of signal to noise ratio for line shape analysis is discussed. Translational (Doppler) temperatures were calculated from the Gaussian components of the line width and were found to be on the order of 6300 0 K. The excitation temperature of iron I was also determined from the same spectral data by the spectroscopic slope method based on the Einstein-Boltzmann expression for spectral intensity and was found to be on the order of 4700 0 K. 31 references

  9. Line roughness improvements on self-aligned quadruple patterning by wafer stress engineering

    Liu, Eric; Ko, Akiteru; Biolsi, Peter; Chae, Soo Doo; Hsieh, Chia-Yun; Kagaya, Munehito; Lee, Choongman; Moriya, Tsuyoshi; Tsujikawa, Shimpei; Suzuki, Yusuke; Okubo, Kazuya; Imai, Kiyotaka

    2018-04-01

    In integrated circuit and memory devices, size shrinkage has been the most effective method to reduce production cost and enable the steady increment of the number of transistors per unit area over the past few decades. In order to reduce the die size and feature size, it is necessary to minimize pattern formation in the advance node development. In the node of sub-10nm, extreme ultra violet lithography (EUV) and multi-patterning solutions based on 193nm immersionlithography are the two most common options to achieve the size requirement. In such small features of line and space pattern, line width roughness (LWR) and line edge roughness (LER) contribute significant amount of process variation that impacts both physical and electrical performances. In this paper, we focus on optimizing the line roughness performance by using wafer stress engineering on 30nm pitch line and space pattern. This pattern is generated by a self-aligned quadruple patterning (SAQP) technique for the potential application of fin formation. Our investigation starts by comparing film materials and stress levels in various processing steps and material selection on SAQP integration scheme. From the cross-matrix comparison, we are able to determine the best stack of film selection and stress combination in order to achieve the lowest line roughness performance while obtaining pattern validity after fin etch. This stack is also used to study the step-by-step line roughness performance from SAQP to fin etch. Finally, we will show a successful patterning of 30nm pitch line and space pattern SAQP scheme with 1nm line roughness performance.

  10. Simulation study of the effect of molar mass dispersity on domain interfacial roughness in lamellae forming block copolymers for directed self-assembly

    Peters, Andrew J; Lawson, Richard A; Nation, Benjamin D; Ludovice, Peter J; Henderson, Clifford L

    2015-01-01

    A coarse-grained molecular dynamics model was used to study the thin film self-assembly and resulting pattern properties of block copolymer (BCP) systems with various molar mass dispersities. Diblock copolymers (i.e. A–b–B type) were simulated in an aligned lamellar state, which is one of the most common patterns of potential use for integrated circuit fabrication via directed self-assembly of BCPs. Effects of the molar mass dispersity (Ð) on feature pitch and interfacial roughness, which are critical lithographic parameters that have a direct impact on integrated circuit performance, were simulated. It was found that for a realistic distribution of polymer molecular weights, modeled by a Wesslau distribution, both line edge roughness (LER) and line width roughness (LWR) increase approximately linearly with increasing Ð, up to ∼45% of the monodisperse value at Ð = 1.5. Mechanisms of compensation for increased A–A and B–B roughness were considered. It was found that long and short chain positions were not correlated, and that long chains were significantly deformed in shape. The increase in LWR was due to the increase in LER and a constant correlation between the line edges. Unaligned systems show a correlation between domain width and local molecular weight, while systems aligned on an alternating pattern of A and B lines did not show any correlation. When the volume fraction of individual chains was allowed to vary, similar results were found when considering the Ð of the block as opposed to the Ð of the entire system. (paper)

  11. 1H line width dependence on MAS speed in solid state NMR - Comparison of experiment and simulation

    Sternberg, Ulrich; Witter, Raiker; Kuprov, Ilya; Lamley, Jonathan M.; Oss, Andres; Lewandowski, Józef R.; Samoson, Ago

    2018-06-01

    Recent developments in magic angle spinning (MAS) technology permit spinning frequencies of ≥100 kHz. We examine the effect of such fast MAS rates upon nuclear magnetic resonance proton line widths in the multi-spin system of β-Asp-Ala crystal. We perform powder pattern simulations employing Fokker-Plank approach with periodic boundary conditions and 1H-chemical shift tensors calculated using the bond polarization theory. The theoretical predictions mirror well the experimental results. Both approaches demonstrate that homogeneous broadening has a linear-quadratic dependency on the inverse of the MAS spinning frequency and that, at the faster end of the spinning frequencies, the residual spectral line broadening becomes dominated by chemical shift distributions and susceptibility effects even for crystalline systems.

  12. Low leaching and low LWR photoresist development for 193 nm immersion lithography

    Ando, Nobuo; Lee, Youngjoon; Miyagawa, Takayuki; Edamatsu, Kunishige; Takemoto, Ichiki; Yamamoto, Satoshi; Tsuchida, Yoshinobu; Yamamoto, Keiko; Konishi, Shinji; Nakano, Katsushi; Tomoharu, Fujiwara

    2006-03-01

    receding contact angle become very important issue for not only defectivity but also scanner throughput. Some of our experimental results along this line of study are also included in the report. The last topic covered is LWR (Line Width Roughness) as an essential leverage for performance improvement, especially for the smaller CD that immersion lithography is aiming to define. Our recent effort to find effect and working concept to reduce LWR with low leaching materials is also described.

  13. Amide I SFG Spectral Line Width Probes the Lipid-Peptide and Peptide-Peptide Interactions at Cell Membrane In Situ and in Real Time.

    Zhang, Baixiong; Tan, Junjun; Li, Chuanzhao; Zhang, Jiahui; Ye, Shuji

    2018-06-13

    The balance of lipid-peptide and peptide-peptide interactions at cell membrane is essential to a large variety of cellular processes. In this study, we have experimentally demonstrated for the first time that sum frequency generation vibrational spectroscopy can be used to probe the peptide-peptide and lipid-peptide interactions in cell membrane in situ and in real time by determination of the line width of amide I band of protein backbone. Using a "benchmark" model of α-helical WALP23, it is found that the dominated lipid-peptide interaction causes a narrow line width of the amide I band, whereas the peptide-peptide interaction can markedly broaden the line width. When WALP23 molecules insert into the lipid bilayer, a quite narrow line width of the amide I band is observed because of the lipid-peptide interaction. In contrast, when the peptide lies down on the bilayer surface, the line width of amide I band becomes very broad owing to the peptide-peptide interaction. In terms of the real-time change in the line width, the transition from peptide-peptide interaction to lipid-peptide interaction is monitored during the insertion of WALP23 into 1,2-dipalmitoyl- sn-glycero-3-phospho-(1'- rac-glycerol) (DPPG) lipid bilayer. The dephasing time of a pure α-helical WALP23 in 1-palmitoyl-2-oleoyl- sn-glycero-3-phospho-(1'- rac-glycerol) and DPPG bilayer is determined to be 2.2 and 0.64 ps, respectively. The peptide-peptide interaction can largely accelerate the dephasing time.

  14. Solar off-limb line widths with SUMER: revised value of the non-thermal velocity and new results

    L. Dolla

    2009-09-01

    Full Text Available Alfvén waves and ion-cyclotron absorption of high-frequency waves are frequently brought into models devoted to coronal heating and fast solar-wind acceleration. Signatures of ion-cyclotron resonance have already been observed in situ in the solar wind and in the upper corona. In the lower corona, one can use the line profiles to infer the ion temperatures. But the value of the so-called "non-thermal" (or "unresolved" velocity, potentially related to the amplitude of Alfvén waves propagating in the corona, is critical in firmly identifying ion-cyclotron preferential heating. In a previous paper, we proposed a method to constrain both the Alfvén wave amplitude and the preferential heating, above a polar coronal hole observed with the SUMER/SOHO spectrometer. Taking into account the effect of instrumental stray light before analysing the line profiles, we ruled out any direct evidence of damping of the Alfvén waves and showed that ions with the lowest charge-to-mass ratios were preferentially heated. We re-analyse these data here to correct the derived non-thermal velocity, and we discuss the consequences on the main results. We also include a measure of the Fe VIII 1442.56 Å line width (second order, thus extending the charge-to-mass ratio domain towards ions more likely to experience cyclotron resonance.

  15. Experimental demonstration of line-width modulation in plasmonic lithography using a solid immersion lens-based active nano-gap control

    Lee, Won-Sup; Kim, Taeseob; Choi, Guk-Jong; Lim, Geon; Joe, Hang-Eun; Gang, Myeong-Gu; Min, Byung-Kwon; Park, No-Cheol; Moon, Hyungbae; Kim, Do-Hyung; Park, Young-Pil

    2015-01-01

    Plasmonic lithography has been used in nanofabrication because of its utility beyond the diffraction limit. The resolution of plasmonic lithography depends on the nano-gap between the nanoaperture and the photoresist surface—changing the gap distance can modulate the line-width of the pattern. In this letter, we demonstrate solid-immersion lens based active non-contact plasmonic lithography, applying a range of gap conditions to modulate the line-width of the pattern. Using a solid-immersion lens-based near-field control system, the nano-gap between the exit surface of the nanoaperture and the media can be actively modulated and maintained to within a few nanometers. The line-widths of the recorded patterns using 15- and 5-nm gaps were 47 and 19.5 nm, respectively, which matched closely the calculated full-width at half-maximum. From these results, we conclude that changing the nano-gap within a solid-immersion lens-based plasmonic head results in varying line-width patterns

  16. White Dwarf Rotation as a Function of Mass and a Dichotomy of Mode Line Widths: Kepler  Observations of 27 Pulsating DA White Dwarfs through K2 Campaign 8

    Hermes, J. J.; Fanale, S. M.; Dennihy, E.; Fuchs, J. T.; Dunlap, B. H.; Clemens, J. C. [Department of Physics and Astronomy, University of North Carolina, Chapel Hill, NC 27599 (United States); Gänsicke, B. T.; Greiss, S.; Tremblay, P.-E.; Fusillo, N. P. Gentile; Raddi, R.; Chote, P.; Marsh, T. R. [Department of Physics, University of Warwick, Coventry CV4 7AL (United Kingdom); Kawaler, Steven D. [Department of Physics and Astronomy, Iowa State University, Ames, IA 50011 (United States); Bell, Keaton J.; Montgomery, M. H.; Winget, D. E. [Department of Astronomy, University of Texas at Austin, Austin, TX 78712 (United States); Redfield, S., E-mail: jjhermes@unc.edu [Wesleyan University Astronomy Department, Van Vleck Observatory, 96 Foss Hill Drive, Middletown, CT 06459 (United States)

    2017-10-01

    We present photometry and spectroscopy for 27 pulsating hydrogen-atmosphere white dwarfs (DAVs; a.k.a. ZZ Ceti stars) observed by the Kepler space telescope up to K2 Campaign 8, an extensive compilation of observations with unprecedented duration (>75 days) and duty cycle (>90%). The space-based photometry reveals pulsation properties previously inaccessible to ground-based observations. We observe a sharp dichotomy in oscillation mode line widths at roughly 800 s, such that white dwarf pulsations with periods exceeding 800 s have substantially broader mode line widths, more reminiscent of a damped harmonic oscillator than a heat-driven pulsator. Extended Kepler coverage also permits extensive mode identification: we identify the spherical degree of 87 out of 201 unique radial orders, providing direct constraints of the rotation period for 20 of these 27 DAVs, more than doubling the number of white dwarfs with rotation periods determined via asteroseismology. We also obtain spectroscopy from 4 m-class telescopes for all DAVs with Kepler photometry. Using these homogeneously analyzed spectra, we estimate the overall mass of all 27 DAVs, which allows us to measure white dwarf rotation as a function of mass, constraining the endpoints of angular momentum in low- and intermediate-mass stars. We find that 0.51–0.73 M {sub ⊙} white dwarfs, which evolved from 1.7–3.0 M {sub ⊙} ZAMS progenitors, have a mean rotation period of 35 hr with a standard deviation of 28 hr, with notable exceptions for higher-mass white dwarfs. Finally, we announce an online repository for our Kepler data and follow-up spectroscopy, which we collect at http://k2wd.org.

  17. The KMOS3D Survey: Rotating Compact Star-forming Galaxies and the Decomposition of Integrated Line Widths

    Wisnioski, E.; Mendel, J. T.; Förster Schreiber, N. M.; Genzel, R.; Wilman, D.; Wuyts, S.; Belli, S.; Beifiori, A.; Bender, R.; Brammer, G.; Chan, J.; Davies, R. I.; Davies, R. L.; Fabricius, M.; Fossati, M.; Galametz, A.; Lang, P.; Lutz, D.; Nelson, E. J.; Momcheva, I.; Rosario, D.; Saglia, R.; Tacconi, L. J.; Tadaki, K.; Übler, H.; van Dokkum, P. G.

    2018-03-01

    Using integral field spectroscopy, we investigate the kinematic properties of 35 massive centrally dense and compact star-forming galaxies (SFGs; {log}{\\overline{M}}* [{M}ȯ ]=11.1, {log}({{{Σ }}}1{kpc}[{M}ȯ {kpc}}-2])> 9.5, {log}({M}* /{r}e1.5[{M}ȯ {kpc}}-1.5])> 10.3) at z ∼ 0.7–3.7 within the KMOS3D survey. We spatially resolve 23 compact SFGs and find that the majority are dominated by rotational motions with velocities ranging from 95 to 500 km s‑1. The range of rotation velocities is reflected in a similar range of integrated Hα line widths, 75–400 km s‑1, consistent with the kinematic properties of mass-matched extended galaxies from the full KMOS3D sample. The fraction of compact SFGs that are classified as “rotation-dominated” or “disklike” also mirrors the fractions of the full KMOS3D sample. We show that integrated line-of-sight gas velocity dispersions from KMOS3D are best approximated by a linear combination of their rotation and turbulent velocities with a lesser but still significant contribution from galactic-scale winds. The Hα exponential disk sizes of compact SFGs are, on average, 2.5 ± 0.2 kpc, 1–2× the continuum sizes, in agreement with previous work. The compact SFGs have a 1.4× higher active galactic nucleus (AGN) incidence than the full KMOS3D sample at fixed stellar mass with an average AGN fraction of 76%. Given their high and centrally concentrated stellar masses, as well as stellar-to-dynamical mass ratios close to unity, the compact SFGs are likely to have low molecular gas fractions and to quench on a short timescale unless replenished with inflowing gas. The rotation in these compact systems suggests that their direct descendants are rotating passive galaxies. Based on observations obtained at the Very Large Telescope (VLT) of the European Southern Observatory (ESO), Paranal, Chile (ESO program IDs 092A-0091, 093.A-0079, 094.A-0217, 095.A-0047, 096.A-0025, 097.A-0028, and 098.A-0045).

  18. Narrow line-width Tm3+ doped double-clad silica fiber laser based on in-line cascade biconical tapers filter

    Tian, Y; Zhao, J Q; Wang, W; Wang, Y Z; Gao, W

    2010-01-01

    Narrow line-width 793 nm laser diode cladding pumped Tm 3+ doped double cladding silica fiber laser with in-line four concatenated tapers filter was reported for the first time to our knowledge. These cascade tapers located 3.6 cm from the output end of the fiber laser was fabricated by heating and stretching method. The taper's transmitted power response as a function of wavelength was described by using local mode coupling theory and successive tapers filter model. The wavelength filter function of the in-line cascade tapers in a linear cavity fiber laser was demonstrated, and the experimental result agreed with these theories. The maximum output laser power was 736 mW, corresponding to single peak of laser spectrum with narrow line-width of ∼ 60 pm

  19. Light Water Reactor (LWR) safety

    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

  20. LWR-core behaviour project

    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.)

  1. A tunable narrow-line-width multi-wavelength Er-doped fiber laser based on a high birefringence fiber ring mirror and an auto-tracking filter

    Jia, Xiu-jie; Liu, Yan-ge; Si, Li-bin; Guo, Zhan-cheng; Fu, Sheng-gui; Kai, Gui-yun; Dong, Xiao-yi

    2008-01-01

    A novel multi-wavelength erbium-doped fiber laser operating in C-band is proposed and successfully demonstrated. The wavelength interval between the wavelengths is about 0.22 nm. The 3 dB bandwidth of the laser is about 0.012 nm, and the output power reaches 4.8 mW. By using a high birefringence fiber ring mirror (HiBi-FLM) and a tunable FBG, the laser realizes switchable and tunable characteristic. The mode hopping can be effectively prevented. Moreover, this laser can improve wavelength stability significantly by taking advantage of an un-pumped Er3+-doped fiber at the standing-wave section. The laser can operate in stable narrow-line-width with single-, dual-wavelength, and unstable triple-wavelength output at room temperature.

  2. Correlation between the line width and the line flux of the double-peaked broad Hα of 3C390.3

    Zhang, Xue-Guang

    2013-03-01

    In this paper, we carefully check the correlation between the line width (second moment) and the line flux of the double-peaked broad Hα of the well-known mapped active galactic nucleus (AGN) 3C390.3 in order to show some further distinctions between double-peaked emitters and normal broad-line AGN. Based on the virialization assumption MBH ∝ RBLR × V2(BLR) and the empirical relation RBLR ∝ L˜0.5, one strong negative correlation between the line width and the line flux of the double-peaked broad lines should be expected for 3C390.3, such as the negative correlation confirmed for the mapped broad-line object NGC 5548, RBLR × V2(BLR) ∝ L˜0.5 × σ2 = constant. Moreover, based on the public spectra around 1995 from the AGN WATCH project for 3C390.3, one reliable positive correlation is found between the line width and the line flux of the double-peaked broad Hα. In the context of the proposed theoretical accretion disc model for double-peaked emitters, the unexpected positive correlation can be naturally explained, due to different time delays for the inner and outer parts of the disc-like broad-line region (BLR) of 3C390.3. Moreover, the virialization assumption is checked and found to be still available for 3C390.3. However, the time-varying size of the BLR of 3C390.3 cannot be expected by the empirical relation RBLR ∝ L˜0.5. In other words, the mean size of the BLR of 3C390.3 can be estimated by the continuum luminosity (line luminosity), while the continuum emission strengthening leads to the size of BLR decreasing (not increasing) in different moments for 3C390.3. Then, we compared our results of 3C390.3 with the previous results reported in the literature for the other double-peaked emitters, and found that before to clearly correct the effects from disc physical parameters varying (such as the effects of disc precession) for long-term observed line spectra, it is not so meaningful to discuss the correlation of the line parameters of double

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

    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

  4. Recycling U and Pu in LWR

    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

  5. LWR physics in SKODA Works

    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)

  6. Rough Finite State Automata and Rough Languages

    Arulprakasam, R.; Perumal, R.; Radhakrishnan, M.; Dare, V. R.

    2018-04-01

    Sumita Basu [1, 2] recently introduced the concept of a rough finite state (semi)automaton, rough grammar and rough languages. Motivated by the work of [1, 2], in this paper, we investigate some closure properties of rough regular languages and establish the equivalence between the classes of rough languages generated by rough grammar and the classes of rough regular languages accepted by rough finite automaton.

  7. Outline of Swedish activities on LWR fuel

    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.

  8. Fatigue management considering LWR coolant environments

    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)

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

    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

  10. Development of training simulator for LWR

    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)

  11. 'CANDLE' burnup regime after LWR regime

    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)

  12. Modeling the economic consequences of LWR accidents

    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

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

    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

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

    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)

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

    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

  16. C IV LINE-WIDTH ANOMALIES

    Denney, Kelly Dianne; Pogge, R.W.; Assef, R.J.

    2013-01-01

    Comparison of six high-redshift quasar spectra obtained with the Large Binocular Telescope with previous observations from the Sloan Digital Sky Survey shows that failure to correctly identify absorption and other problems with accurate characterization of the CIV emission line profile in low S/N...

  17. Creep damage in zircaloy-4 at LWR temperatures

    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. Is it the end of history for LWR safety?

    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)

  19. Recycle of LWR actinides to an IFR

    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.)

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

    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)

  1. Safety research for LWR type reactors

    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

  2. LWR nuclear power plant component failures

    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

  3. Expert system for estimating LWR plutonium production

    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

  4. Problems associated with domestic LWR technology development

    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.)

  5. Perspectives on the economic risks of LWR accidents

    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

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

    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

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

    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

  8. Measurement of surface roughness

    De Chiffre, Leonardo

    This document is used in connection with two 3 hours laboratory exercises that are part of the course GEOMETRICAL METROLOGY AND MACHINE TESTING. The laboratories include a demonstration of the function of roughness measuring instruments plus a series of exercises illustrating roughness measurement...

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

    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

  10. Rough multiple objective decision making

    Xu, Jiuping

    2011-01-01

    Rough Set TheoryBasic concepts and properties of rough sets Rough Membership Rough Intervals Rough FunctionApplications of Rough SetsMultiple Objective Rough Decision Making Reverse Logistics Problem with Rough Interval Parameters MODM based Rough Approximation for Feasible RegionEVRMCCRMDCRM Reverse Logistics Network Design Problem of Suji Renewable Resource MarketBilevel Multiple Objective Rough Decision Making Hierarchical Supply Chain Planning Problem with Rough Interval Parameters Bilevel Decision Making ModelBL-EVRM BL-CCRMBL-DCRMApplication to Supply Chain Planning of Mianyang Co., LtdStochastic Multiple Objective Rough Decision Multi-Objective Resource-Constrained Project Scheduling UnderRough Random EnvironmentRandom Variable Stochastic EVRM Stochastic CCRM Stochastic DCRM Multi-Objective rc-PSP/mM/Ro-Ra for Longtan Hydropower StationFuzzy Multiple Objective Rough Decision Making Allocation Problem under Fuzzy Environment Fuzzy Variable Fu-EVRM Fu-CCRM Fu-DCRM Earth-Rock Work Allocation Problem.

  11. Utility requirements for advanced LWR passive plants

    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.)

  12. Criticality impacts on LWR fuel storage efficiency

    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

  13. Economic analyses of LWR fuel cycles

    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

  14. NUPEC proves reliability of LWR fuel assemblies

    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.)

  15. Advanced LWR Nuclear Fuel Cladding Development

    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)

  16. LWR primary coolant pipe rupture test rig

    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.)

  17. Results of LWR core transient benchmarks

    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

  18. Environmental development plan. LWR commercial waste management

    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

  19. Design consideration on severe accident for future LWR

    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

  20. The minimum attention plant inherent safety through LWR simplification

    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

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

    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

  2. Metal Matrix Microencapsulated Fuel Technology for LWR Applications

    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.

  3. Generalized rough sets

    Rady, E.A.; Kozae, A.M.; Abd El-Monsef, M.M.E.

    2004-01-01

    The process of analyzing data under uncertainty is a main goal for many real life problems. Statistical analysis for such data is an interested area for research. The aim of this paper is to introduce a new method concerning the generalization and modification of the rough set theory introduced early by Pawlak [Int. J. Comput. Inform. Sci. 11 (1982) 314

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

    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

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

    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)

  6. Nondestructive evaluation of LWR spent fuel shipping casks

    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

  7. Materials choices for the advanced LWR steam generators

    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

  8. Contributions to LWR spent fuel storage and transport

    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

  9. Roughing up Beta

    Bollerslev, Tim; Li, Sophia Zhengzi; Todorov, Viktor

    -section. An investment strategy that goes long stocks with high jump betas and short stocks with low jump betas produces significant average excess returns. These higher risk premiums for the discontinuous and overnight market betas remain significant after controlling for a long list of other firm characteristics......Motivated by the implications from a stylized equilibrium pricing framework, we investigate empirically how individual equity prices respond to continuous, or \\smooth," and jumpy, or \\rough," market price moves, and how these different market price risks, or betas, are priced in the cross......-section of expected returns. Based on a novel highfrequency dataset of almost one-thousand individual stocks over two decades, we find that the two rough betas associated with intraday discontinuous and overnight returns entail significant risk premiums, while the intraday continuous beta is not priced in the cross...

  10. Performance and reliability of LWR fuel

    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

  11. Modelling dynamic roughness during floods

    Paarlberg, Andries; Dohmen-Janssen, Catarine M.; Hulscher, Suzanne J.M.H.; Termes, A.P.P.

    2007-01-01

    In this paper, we present a dynamic roughness model to predict water levels during floods. Hysteresis effects of dune development are explicitly included. It is shown that differences between the new dynamic roughness model, and models where the roughness coefficient is calibrated, are most

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

    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)

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

    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

  14. Development of top nozzle for Korean standard LWR fuel

    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.

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

    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

  16. Development of LWR fuel performance code FEMAXI-6

    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.)

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

    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)

  18. Thermal conductivity of heterogeneous LWR MOX fuels

    Staicu, D.; Barker, M.

    2013-11-01

    It is generally observed that the thermal conductivity of LWR MOX fuel is lower than that of pure UO2. For MOX, the degradation is usually only interpreted as an effect of the substitution of U atoms by Pu. This hypothesis is however in contradiction with the observations of Duriez and Philiponneau showing that the thermal conductivity of MOX is independent of the Pu content in the ranges 3-15 and 15-30 wt.% PuO2 respectively. Attributing this degradation to Pu only implies that stoichiometric heterogeneous MOX can be obtained, while we show that any heterogeneity in the plutonium distribution in the sample introduces a variation in the local stoichiometry which in turn has a strong impact on the thermal conductivity. A model quantifying this effect is obtained and a new set of experimental results for homogeneous and heterogeneous MOX fuels is presented and used to validate the proposed model. In irradiated fuels, this effect is predicted to disappear early during irradiation. The 3, 6 and 10 wt.% Pu samples have a similar thermal conductivity. Comparison of the results for this homogeneous microstructure with MIMAS (heterogeneous) fuel of the same composition showed no difference for the Pu contents of 3, 5.9, 6, 7.87 and 10 wt.%. A small increase of the thermal conductivity was obtained for 15 wt.% Pu. This increase is of about 6% when compared to the average of the values obtained for 3, 6 and 10 wt.% Pu. For comparison purposes, Duriez also measured the thermal conductivity of FBR MOX with 21.4 wt.% Pu with O/M = 1.982 and a density close to 95% TD and found a value in good agreement with the estimation obtained using the formula of Philipponneau [8] for FBR MOX, and significantly lower than his results corresponding to the range 3-15 wt.% Pu. This difference in thermal conductivity is of about 20%, i.e. higher than the measurement uncertainties.Thus, a significant difference was observed between FBR and PWR MOX fuels, but was not explained. This difference

  19. Notions of Rough Neutrosophic Digraphs

    Nabeela Ishfaq

    2018-01-01

    Full Text Available [-3]Graph theory has numerous applications in various disciplines, including computer networks, neural networks, expert systems, cluster analysis, and image capturing. Rough neutrosophic set (NS theory is a hybrid tool for handling uncertain information that exists in real life. In this research paper, we apply the concept of rough NS theory to graphs and present a new kind of graph structure, rough neutrosophic digraphs. We present certain operations, including lexicographic products, strong products, rejection and tensor products on rough neutrosophic digraphs. We investigate some of their properties. We also present an application of a rough neutrosophic digraph in decision-making.

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

    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

  1. Safety-related LWR research. Annual report 1993

    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

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

    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

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

    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.

  4. Standard casks for the transport of LWR spent fuel

    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

  5. Investigation of valve failure problems in LWR power plants

    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

  6. Modular approach to LWR in-core fuel management

    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)

  7. Development of information management system on LWR spent fuel

    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

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

    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.

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

    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

  10. Energy profit ratio on LWR by uranium recycles

    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)

  11. Development of information management system on LWR spent fuel

    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.

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

    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.

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

    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

  14. Rough Surface Contact

    T Nguyen

    2017-06-01

    Full Text Available This paper studies the contact of general rough curved surfaces having nearly identical geometries, assuming the contact at each differential area obeys the model proposed by Greenwood and Williamson. In order to account for the most general gross geometry, principles of differential geometry of surface are applied. This method while requires more rigorous mathematical manipulations, the fact that it preserves the original surface geometries thus makes the modeling procedure much more intuitive. For subsequent use, differential geometry of axis-symmetric surface is considered instead of general surface (although this “general case” can be done as well in Chapter 3.1. The final formulas for contact area, load, and frictional torque are derived in Chapter 3.2.

  15. The line roughness improvement with plasma coating and cure treatment for 193nm lithography and beyond

    Zheng, Erhu; Huang, Yi; Zhang, Haiyang

    2017-03-01

    As CMOS technology reaches 14nm node and beyond, one of the key challenges of the extension of 193nm immersion lithography is how to control the line edge and width roughness (LER/LWR). For Self-aligned Multiple Patterning (SaMP), LER becomes larger while LWR becomes smaller as the process proceeds[1]. It means plasma etch process becomes more and more dominant for LER reduction. In this work, we mainly focus on the core etch solution including an extra plasma coating process introduced before the bottom anti reflective coating (BARC) open step, and an extra plasma cure process applied right after BARC-open step. Firstly, we leveraged the optimal design experiment (ODE) to investigate the impact of plasma coating step on LER and identified the optimal condition. ODE is an appropriate method for the screening experiments of non-linear parameters in dynamic process models, especially for high-cost-intensive industry [2]. Finally, we obtained the proper plasma coating treatment condition that has been proven to achieve 32% LER improvement compared with standard process. Furthermore, the plasma cure scheme has been also optimized with ODE method to cover the LWR degradation induced by plasma coating treatment.

  16. Roughness Effects on Fretting Fatigue

    Yue, Tongyan; Abdel Wahab, Magd

    2017-05-01

    Fretting is a small oscillatory relative motion between two normal loaded contact surfaces. It may cause fretting fatigue, fretting wear and/or fretting corrosion damage depending on various fretting couples and working conditions. Fretting fatigue usually occurs at partial slip condition, and results in catastrophic failure at the stress levels below the fatigue limit of the material. Many parameters may affect fretting behaviour, including the applied normal load and displacement, material properties, roughness of the contact surfaces, frequency, etc. Since fretting damage is undesirable due to contacting, the effect of rough contact surfaces on fretting damage has been studied by many researchers. Experimental method on this topic is usually focusing on rough surface effects by finishing treatment and random rough surface effects in order to increase fretting fatigue life. However, most of numerical models on roughness are based on random surface. This paper reviewed both experimental and numerical methodology on the rough surface effects on fretting fatigue.

  17. Variations in roughness predictions (flume experiments)

    Noordam, Daniëlle; Blom, Astrid; van der Klis, H.; Hulscher, Suzanne J.M.H.; Makaske, A.; Wolfert, H.P.; van Os, A.G.

    2005-01-01

    Data of flume experiments with bed forms are used to analyze and compare different roughness predictors. In this study, the hydraulic roughness consists of grain roughness and form roughness. We predict the grain roughness by means of the size of the sediment. The form roughness is predicted by

  18. Comparison of vegetation roughness descriptions

    Augustijn, Dionysius C.M.; Huthoff, Freek; van Velzen, E.H.; Altinakar, M.S.; Kokpinar, M.A.; Aydin, I.; Cokgor, S.; Kirkgoz, S.

    2008-01-01

    Vegetation roughness is an important parameter in describing flow through river systems. Vegetation impedes the flow, which affects the stage-discharge curve and may increase flood risks. Roughness is often used as a calibration parameter in river models, however when vegetation is allowed to

  19. Bankruptcy Prediction with Rough Sets

    J.C. Bioch (Cor); V. Popova (Viara)

    2001-01-01

    textabstractThe bankruptcy prediction problem can be considered an or dinal classification problem. The classical theory of Rough Sets describes objects by discrete attributes, and does not take into account the order- ing of the attributes values. This paper proposes a modification of the Rough Set

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

    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

  1. Conceptual design of a spent LWR fuel recycle complex

    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

  2. Issues in risk analysis of passive LWR designs

    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

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

    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

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

    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)

  5. Nondestructive evaluation of LWR spent fuel shipping casks

    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

  6. Standard casks for the transport of LWR spent fuel

    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

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

    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

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

    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

  9. Investigation of valve failure problems in LWR power plants

    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%).

  10. Transmutation of LWR waste actinides in thermal reactors

    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

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

    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

  12. Pie technique of LWR fuel cladding fracture toughness test

    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)

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

    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)

  14. Generalizing roughness: experiments with flow-oriented roughness

    Trevisani, Sebastiano

    2015-04-01

    Surface texture analysis applied to High Resolution Digital Terrain Models (HRDTMs) improves the capability to characterize fine-scale morphology and permits the derivation of useful morphometric indexes. An important indicator to be taken into account in surface texture analysis is surface roughness, which can have a discriminant role in the detection of different geomorphic processes and factors. The evaluation of surface roughness is generally performed considering it as an isotropic surface parameter (e.g., Cavalli, 2008; Grohmann, 2011). However, surface texture has often an anisotropic character, which means that surface roughness could change according to the considered direction. In some applications, for example involving surface flow processes, the anisotropy of roughness should be taken into account (e.g., Trevisani, 2012; Smith, 2014). Accordingly, we test the application of a flow-oriented directional measure of roughness, computed considering surface gravity-driven flow. For the calculation of flow-oriented roughness we use both classical variogram-based roughness (e.g., Herzfeld,1996; Atkinson, 2000) as well as an ad-hoc developed robust modification of variogram (i.e. MAD, Trevisani, 2014). The presented approach, based on a D8 algorithm, shows the potential impact of considering directionality in the calculation of roughness indexes. The use of flow-oriented roughness could improve the definition of effective proxies of impedance to flow. Preliminary results on the integration of directional roughness operators with morphometric-based models, are promising and can be extended to more complex approaches. Atkinson, P.M., Lewis, P., 2000. Geostatistical classification for remote sensing: an introduction. Computers & Geosciences 26, 361-371. Cavalli, M. & Marchi, L. 2008, "Characterization of the surface morphology of an alpine alluvial fan using airborne LiDAR", Natural Hazards and Earth System Science, vol. 8, no. 2, pp. 323-333. Grohmann, C

  15. The dupic fuel cycle synergism between LWR and HWR

    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)

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

    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)

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

    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

  18. Qualification of ARROTTA code for LWR accident analysis

    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)

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

    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.

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

    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.)

  1. Armor Plate Surface Roughness Measurements

    Stanton, Brian; Coburn, William; Pizzillo, Thomas J

    2005-01-01

    ...., surface texture and coatings) that could become important at high frequency. We measure waviness and roughness of various plates to know the parameter range for smooth aluminum and rolled homogenous armor (RHA...

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

    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)

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

    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

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

    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

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

    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.)

  6. Alternatives for managing post LWR reactor nuclear wastes

    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

  7. LWR-PV Surveillance Dosimetry Improvement Program review graphics

    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

  8. LWR risk management by safety R and D

    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

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

    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

  10. Evaluation of management alternatives for LWR hulls and caps

    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

  11. Cost optimization of long-cycle LWR operation

    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

  12. Irradiation effects on thermal properties of LWR hydride fuel

    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.

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

    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

  14. Fracture probability evaluation of a LWR pressure vessel

    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

  15. Dual-purpose LWR supplying heat for desalination

    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.)

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

    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

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

    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)

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

    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)

  19. Progress in Development of I2S-LWR Concept

    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)

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

    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.)

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

    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)

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

    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

  3. Review of the margins for ASME code fatigue design curve - effects of surface roughness and material variability

    Chopra, O. K.; Shack, W. J.

    2003-01-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. The Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ((var e psilon)-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of the existing fatigue (var e psilon)-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue (var e psilon)-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for ASME Code fatigue design curves is presented

  4. Review of the margins for ASME code fatigue design curve - effects of surface roughness and material variability.

    Chopra, O. K.; Shack, W. J.; Energy Technology

    2003-10-03

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. The Code specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain-vs.-life ({var_epsilon}-N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This report provides an overview of the existing fatigue {var_epsilon}-N data for carbon and low-alloy steels and wrought and cast austenitic SSs to define the effects of key material, loading, and environmental parameters on the fatigue lives of the steels. Experimental data are presented on the effects of surface roughness on the fatigue life of these steels in air and LWR environments. Statistical models are presented for estimating the fatigue {var_epsilon}-N curves as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are discussed. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue evaluations. A critical review of the margins for ASME Code fatigue design curves is presented.

  5. Stochastic control with rough paths

    Diehl, Joscha; Friz, Peter K.; Gassiat, Paul

    2017-01-01

    We study a class of controlled differential equations driven by rough paths (or rough path realizations of Brownian motion) in the sense of Lyons. It is shown that the value function satisfies a HJB type equation; we also establish a form of the Pontryagin maximum principle. Deterministic problems of this type arise in the duality theory for controlled diffusion processes and typically involve anticipating stochastic analysis. We make the link to old work of Davis and Burstein (Stoch Stoch Rep 40:203–256, 1992) and then prove a continuous-time generalization of Roger’s duality formula [SIAM J Control Optim 46:1116–1132, 2007]. The generic case of controlled volatility is seen to give trivial duality bounds, and explains the focus in Burstein–Davis’ (and this) work on controlled drift. Our study of controlled rough differential equations also relates to work of Mazliak and Nourdin (Stoch Dyn 08:23, 2008).

  6. Heat transfer from rough surfaces

    Dalle Donne, M.

    1977-01-01

    Artificial roughness is often used in nuclear reactors to improve the thermal performance of the fuel elements. Although these are made up of clusters of rods, the experiments to measure the heat transfer and friction coefficients of roughness are performed with single rods contained in smooth tubes. This work illustrated a new transformation method to obtain data applicable to reactor fuel elements from these annulus experiments. New experimental friction data are presented for ten rods, each with a different artificial roughness made up of two-dimensional rectangular ribs. For each rod four tests have been performed, each in a different outer smooth tube. For two of these rods, each for two different outer tubes, heat transfer data are also given. The friction and heat transfer data, transformed with the present method, are correlated by simple equations. In the paper, these equations are applied to a case typical for a Gas Cooled Fast Reactor fuel element. (orig.) [de

  7. Stochastic control with rough paths

    Diehl, Joscha [University of California San Diego (United States); Friz, Peter K., E-mail: friz@math.tu-berlin.de [TU & WIAS Berlin (Germany); Gassiat, Paul [CEREMADE, Université Paris-Dauphine, PSL Research University (France)

    2017-04-15

    We study a class of controlled differential equations driven by rough paths (or rough path realizations of Brownian motion) in the sense of Lyons. It is shown that the value function satisfies a HJB type equation; we also establish a form of the Pontryagin maximum principle. Deterministic problems of this type arise in the duality theory for controlled diffusion processes and typically involve anticipating stochastic analysis. We make the link to old work of Davis and Burstein (Stoch Stoch Rep 40:203–256, 1992) and then prove a continuous-time generalization of Roger’s duality formula [SIAM J Control Optim 46:1116–1132, 2007]. The generic case of controlled volatility is seen to give trivial duality bounds, and explains the focus in Burstein–Davis’ (and this) work on controlled drift. Our study of controlled rough differential equations also relates to work of Mazliak and Nourdin (Stoch Dyn 08:23, 2008).

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

    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

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

    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

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

    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.

  11. Information Measures of Roughness of Knowledge and Rough Sets for Incomplete Information Systems

    LIANG Ji-ye; QU Kai-she

    2001-01-01

    In this paper we address information measures of roughness of knowledge and rough sets for incomplete information systems. The definition of rough entropy of knowledge and its important properties are given. In particular, the relationship between rough entropy of knowledge and the Hartley measure of uncertainty is established. We show that rough entropy of knowledge decreases monotonously as granularity of information become smaller. This gives an information interpretation for roughness of knowledge. Based on rough entropy of knowledge and roughness of rough set. a definition of rough entropy of rough set is proposed, and we show that rough entropy of rough set decreases monotonously as granularity of information become smaller. This gives more accurate measure for roughness of rough set.

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

    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

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

    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

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

    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

  15. Does Surface Roughness Amplify Wetting?

    Malijevský, Alexandr

    2014-01-01

    Roč. 141, č. 18 (2014), s. 184703 ISSN 0021-9606 R&D Projects: GA ČR GA13-09914S Institutional support: RVO:67985858 Keywords : density functional theory * wetting * roughness Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 2.952, year: 2014

  16. Calibration of surface roughness standards

    Thalmann, R.; Nicolet, A.; Meli, F.

    2016-01-01

    organisations. Five surface texture standards of different type were circulated and on each of the standards several roughness parameters according to the standard ISO 4287 had to be determined. 32 out of 395 individual results were not consistent with the reference value. After some corrective actions...

  17. Fuzzy Rough Ring and Its Prop erties

    REN Bi-jun; FU Yan-ling

    2013-01-01

    This paper is devoted to the theories of fuzzy rough ring and its properties. The fuzzy approximation space generated by fuzzy ideals and the fuzzy rough approximation operators were proposed in the frame of fuzzy rough set model. The basic properties of fuzzy rough approximation operators were analyzed and the consistency between approximation operators and the binary operation of ring was discussed.

  18. Human roughness perception and possible factors effecting roughness sensation.

    Aktar, Tugba; Chen, Jianshe; Ettelaie, Rammile; Holmes, Melvin; Henson, Brian

    2017-06-01

    Surface texture sensation is significant for business success, in particular for solid surfaces for most of the materials; including foods. Mechanisms of roughness perception are still unknown, especially under different conditions such as lubricants with varying viscosities, different temperatures, or under different force loads during the observation of the surface. This work aims to determine the effect of those unknown factors, with applied sensory tests on 62 healthy participants. Roughness sensation of fingertip was tested under different lubricants including water and diluted syrup solutions at room temperature (25C) and body temperature (37C) by using simple pair-wise comparison to observe the just noticeable difference threshold and perception levels. Additionally, in this research applied force load during roughness observation was tested with pair-wise ranking method to illustrate its possible effect on human sensation. Obtained results showed that human's capability of roughness discrimination reduces with increased viscosity of the lubricant, where the influence of the temperature was not found to be significant. Moreover, the increase in the applied force load showed an increase in the sensitivity of roughness discrimination. Observed effects of the applied factors were also used for estimating the oral sensation of texture during eating. These findings are significant for our fundamental understanding to texture perception, and for the development of new food products with controlled textural features. Texture discrimination ability, more specifically roughness discrimination capability, is a significant factor for preference and appreciation for a wide range of materials, including food, furniture, or fabric. To explore the mechanism of sensation capability through tactile senses, it is necessary to identify the relevant factors and define characteristics that dominate the process involved. The results that will be obtained under these principles

  19. Behavior of LWR fuel elements under accident conditions

    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

  20. Multiscale Analysis of the Roughness Effect on Lubricated Rough Contact

    Demirci , Ibrahim; MEZGHANI , Sabeur; YOUSFI , Mohammed; El Mansori , Mohamed

    2014-01-01

    Determining friction is as equally essential as determining the film thickness in the lubricated contact, and is an important research subject. Indeed, reduction of friction in the automotive industry is important for both the minimization of fuel consumption as well as the decrease in the emissions of greenhouse gases. However, the progress in friction reduction has been limited by the difficulty in understanding the mechanism of roughness effects on friction. It was observed that micro-surf...

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

    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)

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

    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

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

    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

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

    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

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

    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

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

    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

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

    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

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

    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)

  9. Research on ultrasonic flow detection techniques for LWR facilities

    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.)

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

    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)

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

    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

  12. The scale analysis sequence for LWR fuel depletion

    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

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

    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.)

  14. Preliminary concepts for detecting diversion of LWR spent fuel

    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

  15. Discussion of OECD LWR Uncertainty Analysis in Modelling Benchmark

    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)

  16. Ultrasonic backward radiation on painted rough interface

    Kwon, Yong Gyu; Yoon, Seok Soo; Kwon, Sung Duck

    2002-01-01

    The angular dependence(profile) of backscattered ultrasound was measured for steel and brass specimens with periodical surface roughness (1-71μm). Backward radiations showed more linear dependency than normal profile. Direct amplitude increased and averaging amplitude decreased with surface roughness. Painting treatment improved the linearity in direct backward radiation below roughness of 0.03. Scholte and Rayleigh-like waves were observed in the spectrum of averaging backward radiation on periodically rough surface. Painting on periodically rough surface could be used in removing the interface mode effect by periodic roughness.

  17. Enhanced Accident Tolerant LWR Fuels National Metrics Workshop Report

    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

  18. Towards predictive models for transitionally rough surfaces

    Abderrahaman-Elena, Nabil; Garcia-Mayoral, Ricardo

    2017-11-01

    We analyze and model the previously presented decomposition for flow variables in DNS of turbulence over transitionally rough surfaces. The flow is decomposed into two contributions: one produced by the overlying turbulence, which has no footprint of the surface texture, and one induced by the roughness, which is essentially the time-averaged flow around the surface obstacles, but modulated in amplitude by the first component. The roughness-induced component closely resembles the laminar steady flow around the roughness elements at the same non-dimensional roughness size. For small - yet transitionally rough - textures, the roughness-free component is essentially the same as over a smooth wall. Based on these findings, we propose predictive models for the onset of the transitionally rough regime. Project supported by the Engineering and Physical Sciences Research Council (EPSRC).

  19. Rough set classification based on quantum logic

    Hassan, Yasser F.

    2017-11-01

    By combining the advantages of quantum computing and soft computing, the paper shows that rough sets can be used with quantum logic for classification and recognition systems. We suggest the new definition of rough set theory as quantum logic theory. Rough approximations are essential elements in rough set theory, the quantum rough set model for set-valued data directly construct set approximation based on a kind of quantum similarity relation which is presented here. Theoretical analyses demonstrate that the new model for quantum rough sets has new type of decision rule with less redundancy which can be used to give accurate classification using principles of quantum superposition and non-linear quantum relations. To our knowledge, this is the first attempt aiming to define rough sets in representation of a quantum rather than logic or sets. The experiments on data-sets have demonstrated that the proposed model is more accuracy than the traditional rough sets in terms of finding optimal classifications.

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

    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.

  1. Survey of LWR environmental control technology performance and cost

    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

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

    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.

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

    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

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

    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

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

    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)

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

    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.

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

    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.

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

    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)

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

    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

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

    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

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

    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

  12. 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

    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

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

    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.

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

    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)

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

    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

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

    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.)

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

    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.

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

    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

  19. Computer simulations of a rough sphere fluid

    Lyklema, J.W.

    1978-01-01

    A computer simulation is described on rough hard spheres with a continuously variable roughness parameter, including the limits of smooth and completely rough spheres. A system of 500 particles is simulated with a homogeneous mass distribution at 8 different densities and for 5 different values of the roughness parameter. For these 40 physically different situations the intermediate scattering function for 6 values of the wave number, the orientational correlation functions and the velocity autocorrelation functions have been calculated. A comparison has been made with a neutron scattering experiment on neopentane and agreement was good for an intermediate value of the roughness parameter. Some often made approximations in neutron scattering experiments are also checked. The influence of the variable roughness parameter on the correlation functions has been investigated and three simple stochastic models studied to describe the orientational correlation function which shows the most pronounced dependence on the roughness. (Auth.)

  20. Axis Problem of Rough 3-Valued Algebras

    Jianhua Dai; Weidong Chen; Yunhe Pan

    2006-01-01

    The collection of all the rough sets of an approximation space has been given several algebraic interpretations, including Stone algebras, regular double Stone algebras, semi-simple Nelson algebras, pre-rough algebras and 3-valued Lukasiewicz algebras. A 3-valued Lukasiewicz algebra is a Stone algebra, a regular double Stone algebra, a semi-simple Nelson algebra, a pre-rough algebra. Thus, we call the algebra constructed by the collection of rough sets of an approximation space a rough 3-valued Lukasiewicz algebra. In this paper,the rough 3-valued Lukasiewicz algebras, which are a special kind of 3-valued Lukasiewicz algebras, are studied. Whether the rough 3-valued Lukasiewicz algebra is a axled 3-valued Lukasiewicz algebra is examined.

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

    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.

  2. LWR safety research in the Federal Republic of Germany

    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

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

    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

  4. Sensing roughness and polish direction

    Jakobsen, Michael Linde; Olesen, Anders Sig; Larsen, Henning Engelbrecht

    2016-01-01

    As a part of the work carried out in a project supported by the Danish Council for Technology and Innovation, we have investigated the option of smoothing standard CNC-machined surfaces. In the process of constructing optical prototypes, involving custom-designed optics, the development cost...... and time consumption can become prohibitive in a research budget. Machining the optical surfaces directly is expensive and time consuming. Alternatively, a more standardized and cheaper machining method can be used, calling for the object to be manually polished. During the polishing process, the operator...... needs information about the RMS-value of the surface roughness and the current direction of the scratches introduced by the polishing process. The RMS-value indicates to the operator how far he is from the final finish, and the scratch orientation is often specified by the customer in order to avoid...

  5. Software testing in roughness calculation

    Chen, Y L; Hsieh, P F; Fu, W E

    2005-01-01

    A test method to determine the function quality provided by the software for roughness measurement is presented in this study. The function quality of the software requirements should be part of and assessed through the entire life cycle of the software package. The specific function, or output accuracy, is crucial for the analysis of the experimental data. For scientific applications, however, commercial software is usually embedded with specific instrument, which is used for measurement or analysis during the manufacture process. In general, the error ratio caused by the software would be more apparent especially when dealing with relatively small quantities, like the measurements in the nanometer-scale range. The model of 'using a data generator' proposed by NPL of UK was applied in this study. An example of the roughness software is tested and analyzed by the above mentioned process. After selecting the 'reference results', the 'reference data' was generated by a programmable 'data generator'. The filter function of 0.8 mm long cutoff value, defined in ISO 11562 was tested with 66 sinusoid data at different wavelengths. Test results from commercial software and CMS written program were compared to the theoretical data calculated from ISO standards. As for the filter function in this software, the result showed a significant disagreement between the reference and test results. The short cutoff feature for filtering at the high frequencies does not function properly, while the long cutoff feature has the maximum difference in the filtering ratio, which is more than 70% between the wavelength of 300 μm and 500 μm. Conclusively, the commercial software needs to be tested more extensively for specific application by appropriate design of reference dataset to ensure its function quality

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

    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

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

    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

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

    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)

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

    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

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

    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.)

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

    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.

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

    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.

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

    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.

  14. Simplified Approach to Predicting Rough Surface Transition

    Boyle, Robert J.; Stripf, Matthias

    2009-01-01

    Turbine vane heat transfer predictions are given for smooth and rough vanes where the experimental data show transition moving forward on the vane as the surface roughness physical height increases. Consiste nt with smooth vane heat transfer, the transition moves forward for a fixed roughness height as the Reynolds number increases. Comparison s are presented with published experimental data. Some of the data ar e for a regular roughness geometry with a range of roughness heights, Reynolds numbers, and inlet turbulence intensities. The approach ta ken in this analysis is to treat the roughness in a statistical sense , consistent with what would be obtained from blades measured after e xposure to actual engine environments. An approach is given to determ ine the equivalent sand grain roughness from the statistics of the re gular geometry. This approach is guided by the experimental data. A roughness transition criterion is developed, and comparisons are made with experimental data over the entire range of experimental test co nditions. Additional comparisons are made with experimental heat tran sfer data, where the roughness geometries are both regular as well a s statistical. Using the developed analysis, heat transfer calculatio ns are presented for the second stage vane of a high pressure turbine at hypothetical engine conditions.

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

    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

  16. Rock discontinuity surface roughness variation with scale

    Bitenc, Maja; Kieffer, D. Scott; Khoshelham, Kourosh

    2017-04-01

    ABSTRACT: Rock discontinuity surface roughness refers to local departures of the discontinuity surface from planarity and is an important factor influencing the shear resistance. In practice, the Joint Roughness Coefficient (JRC) roughness parameter is commonly relied upon and input to a shear strength criterion such as developed by Barton and Choubey [1977]. The estimation of roughness by JRC is hindered firstly by the subjective nature of visually comparing the joint profile to the ten standard profiles. Secondly, when correlating the standard JRC values and other objective measures of roughness, the roughness idealization is limited to a 2D profile of 10 cm length. With the advance of measuring technologies that provide accurate and high resolution 3D data of surface topography on different scales, new 3D roughness parameters have been developed. A desirable parameter is one that describes rock surface geometry as well as the direction and scale dependency of roughness. In this research a 3D roughness parameter developed by Grasselli [2001] and adapted by Tatone and Grasselli [2009] is adopted. It characterizes surface topography as the cumulative distribution of local apparent inclination of asperities with respect to the shear strength (analysis) direction. Thus, the 3D roughness parameter describes the roughness amplitude and anisotropy (direction dependency), but does not capture the scale properties. In different studies the roughness scale-dependency has been attributed to data resolution or size of the surface joint (see a summary of researches in [Tatone and Grasselli, 2012]). Clearly, the lower resolution results in lower roughness. On the other hand, have the investigations of surface size effect produced conflicting results. While some studies have shown a decrease in roughness with increasing discontinuity size (negative scale effect), others have shown the existence of positive scale effects, or both positive and negative scale effects. We

  17. 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

    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

  18. Bed roughness experiments in supply limited conditions

    Spekkers, Matthieu; Tuijnder, Arjan; Ribberink, Jan S.; Hulscher, Suzanne J.M.H.; Parsons, D.R.; Garlan, T.; Best, J.L.

    2008-01-01

    Reliable roughness models are of great importance, for example, when predicting water levels in rivers. The currently available roughness models are based on fully mobile bed conditions. However, in rivers where widely graded sediments are present more or less permanent armour layers can develop

  19. Axiomatic Characterizations of IVF Rough Approximation Operators

    Guangji Yu

    2014-01-01

    Full Text Available This paper is devoted to the study of axiomatic characterizations of IVF rough approximation operators. IVF approximation spaces are investigated. The fact that different IVF operators satisfy some axioms to guarantee the existence of different types of IVF relations which produce the same operators is proved and then IVF rough approximation operators are characterized by axioms.

  20. Wall roughness induces asymptotic ultimate turbulence

    Zhu, Xiaojue; Verschoof, Ruben Adriaan; Bakhuis, Dennis; Huisman, Sander Gerard; Verzicco, Roberto; Sun, Chao; Lohse, Detlef

    2018-01-01

    Turbulence governs the transport of heat, mass and momentum on multiple scales. In real-world applications, wall-bounded turbulence typically involves surfaces that are rough; however, characterizing and understanding the effects of wall roughness on turbulence remains a challenge. Here, by

  1. Electrochemically grown rough-textured nanowires

    Tyagi, Pawan; Postetter, David; Saragnese, Daniel; Papadakis, Stergios J.; Gracias, David H.

    2010-01-01

    Nanowires with a rough surface texture show unusual electronic, optical, and chemical properties; however, there are only a few existing methods for producing these nanowires. Here, we describe two methods for growing both free standing and lithographically patterned gold (Au) nanowires with a rough surface texture. The first strategy is based on the deposition of nanowires from a silver (Ag)-Au plating solution mixture that precipitates an Ag-Au cyanide complex during electrodeposition at low current densities. This complex disperses in the plating solution, thereby altering the nanowire growth to yield a rough surface texture. These nanowires are mass produced in alumina membranes. The second strategy produces long and rough Au nanowires on lithographically patternable nickel edge templates with corrugations formed by partial etching. These rough nanowires can be easily arrayed and integrated with microscale devices.

  2. Modeling surface roughness scattering in metallic nanowires

    Moors, Kristof, E-mail: kristof@itf.fys.kuleuven.be [KU Leuven, Institute for Theoretical Physics, Celestijnenlaan 200D, B-3001 Leuven (Belgium); IMEC, Kapeldreef 75, B-3001 Leuven (Belgium); Sorée, Bart [IMEC, Kapeldreef 75, B-3001 Leuven (Belgium); Physics Department, University of Antwerp, Groenenborgerlaan 171, B-2020 Antwerpen (Belgium); KU Leuven, Electrical Engineering (ESAT) Department, Kasteelpark Arenberg 10, B-3001 Leuven (Belgium); Magnus, Wim [IMEC, Kapeldreef 75, B-3001 Leuven (Belgium); Physics Department, University of Antwerp, Groenenborgerlaan 171, B-2020 Antwerpen (Belgium)

    2015-09-28

    Ando's model provides a rigorous quantum-mechanical framework for electron-surface roughness scattering, based on the detailed roughness structure. We apply this method to metallic nanowires and improve the model introducing surface roughness distribution functions on a finite domain with analytical expressions for the average surface roughness matrix elements. This approach is valid for any roughness size and extends beyond the commonly used Prange-Nee approximation. The resistivity scaling is obtained from the self-consistent relaxation time solution of the Boltzmann transport equation and is compared to Prange-Nee's approach and other known methods. The results show that a substantial drop in resistivity can be obtained for certain diameters by achieving a large momentum gap between Fermi level states with positive and negative momentum in the transport direction.

  3. Suppression of intrinsic roughness in encapsulated graphene

    Thomsen, Joachim Dahl; Gunst, Tue; Gregersen, Søren Schou

    2017-01-01

    Roughness in graphene is known to contribute to scattering effects which lower carrier mobility. Encapsulating graphene in hexagonal boron nitride (hBN) leads to a significant reduction in roughness and has become the de facto standard method for producing high-quality graphene devices. We have...... fabricated graphene samples encapsulated by hBN that are suspended over apertures in a substrate and used noncontact electron diffraction measurements in a transmission electron microscope to measure the roughness of encapsulated graphene inside such structures. We furthermore compare the roughness...... of these samples to suspended bare graphene and suspended graphene on hBN. The suspended heterostructures display a root mean square (rms) roughness down to 12 pm, considerably less than that previously reported for both suspended graphene and graphene on any substrate and identical within experimental error...

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

    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)

  5. Surface roughness effects on turbulent Couette flow

    Lee, Young Mo; Lee, Jae Hwa

    2017-11-01

    Direct numerical simulation of a turbulent Couette flow with two-dimensional (2-D) rod roughness is performed to examine the effects of the surface roughness. The Reynolds number based on the channel centerline laminar velocity (Uco) and channel half height (h) is Re =7200. The 2-D rods are periodically arranged with a streamwise pitch of λ = 8 k on the bottom wall, and the roughness height is k = 0.12 h. It is shown that the wall-normal extent for the logarithmic layer is significantly shortened in the rough-wall turbulent Couette flow, compared to a turbulent Couette flow with smooth wall. Although the Reynolds stresses are increased in a turbulent channel flow with surface roughness in the outer layer due to large-scale ejection motions produced by the 2-D rods, those of the rough-wall Couette flow are decreased. Isosurfaces of the u-structures averaged in time suggest that the decrease of the turbulent activity near the centerline is associated with weakened large-scale counter-rotating roll modes by the surface roughness. This research was supported by the National Research Foundation of Korea (NRF) funded by the Ministry of Education (NRF-2017R1D1A1A09000537) and the Ministry of Science, ICT & Future Planning (NRF-2017R1A5A1015311).

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

    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

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

    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

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

    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

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

    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

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

    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

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

    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

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

    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.

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

    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

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

    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

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

    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.

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

    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)

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

    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

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

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

    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.

  1. Skin friction measurements of systematically-varied roughness: Probing the role of roughness amplitude and skewness

    Barros, Julio; Flack, Karen; Schultz, Michael

    2017-11-01

    Real-world engineering systems which feature either external or internal wall-bounded turbulent flow are routinely affected by surface roughness. This gives rise to performance degradation in the form of increased drag or head loss. However, at present there is no reliable means to predict these performance losses based upon the roughness topography alone. This work takes a systematic approach by generating random surface roughness in which the surface statistics are closely controlled. Skin friction and roughness function results will be presented for two groups of these rough surfaces. The first group is Gaussian (i.e. zero skewness) in which the root-mean-square roughness height (krms) is varied. The second group has a fixed krms, and the skewness is varied from approximately -1 to +1. The effect of the roughness amplitude and skewness on the skin friction will be discussed. Particular attention will be paid to the effect of these parameters on the roughness function in the transitionally-rough flow regime. For example, the role these parameters play in the monotonic or inflectional nature of the roughness function will be addressed. Future research into the details of the turbulence structure over these rough surfaces will also be outlined. Research funded by U.S. Office of Naval Research (ONR).

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

    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.

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

    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

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

    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)

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

    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

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

    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

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

    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.

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

    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

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

    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.

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

    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)

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

    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

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

    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)

  13. Numerical Schemes for Rough Parabolic Equations

    Deya, Aurelien, E-mail: deya@iecn.u-nancy.fr [Universite de Nancy 1, Institut Elie Cartan Nancy (France)

    2012-04-15

    This paper is devoted to the study of numerical approximation schemes for a class of parabolic equations on (0,1) perturbed by a non-linear rough signal. It is the continuation of Deya (Electron. J. Probab. 16:1489-1518, 2011) and Deya et al. (Probab. Theory Relat. Fields, to appear), where the existence and uniqueness of a solution has been established. The approach combines rough paths methods with standard considerations on discretizing stochastic PDEs. The results apply to a geometric 2-rough path, which covers the case of the multidimensional fractional Brownian motion with Hurst index H>1/3.

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

    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

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

    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

  16. APEX nuclear fuel cycle for production of LWR fuel and elimination of radioactive waste

    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

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

    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

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

    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)

  19. Spin Hall effect by surface roughness

    Zhou, Lingjun; Grigoryan, Vahram L.; Maekawa, Sadamichi; Wang, Xuhui; Xiao, Jiang

    2015-01-01

    induced by surface roughness subscribes only to the side-jump contribution but not the skew scattering. The paradigm proposed in this paper provides the second, not if only, alternative to generate a sizable spin Hall effect.

  20. Roughness coefficients for stream channels in Arizona

    Aldridge, B.N.; Garrett, J.M.

    1973-01-01

    When water flows in an open channel, energy is lost through friction along the banks and bed of the channel and through turbulence within the channel. The amount of energy lost is governed by channel roughness, which is expressed in terms of a roughness coefficient. An evaluation of the roughness coefficient is necessary in many hydraulic computations that involve flow in an open channel. Owing to the lack of satisfactory quantitative procedure, the ability of evaluate roughness coefficients can be developed only through experience; however, a basic knowledge of the methods used to assign the coefficients and the factors affecting them will be a great help. One of the most commonly used equations in open-channel hydraulics is that of Manning. The Manning equation is       1.486

  1. Investigation on Surface Roughness in Cylindrical Grinding

    Rudrapati, Ramesh; Bandyopadhyay, Asish; Pal, Pradip Kumar

    2011-01-01

    Cylindrical grinding is a complex machining process. And surface roughness is often a key factor in any machining process while considering the machine tool or machining performance. Further, surface roughness is one of the measures of the technological quality of the product and is a factor that greatly influences cost and quality. The present work is related to some aspects of surface finish in the context of traverse-cut cylindrical grinding. The parameters considered have been: infeed, longitudinal feed and work speed. Taguchi quality design is used to design the experiments and to identify the significantly import parameter(s) affecting the surface roughness. By utilization of Response Surface Methodology (RSM), second order differential equation has been developed and attempts have also been made for optimization of the process in the context of surface roughness by using C- programming.

  2. Rough horizontal plates: heat transfer and hysteresis

    Tisserand, J-C; Gasteuil, Y; Pabiou, H; Castaing, B; Chilla, F [Universite de Lyon, ENS Lyon, CNRS, 46 Allee d' ltalie, 69364 Lyon Cedex 7 (France); Creyssels, M [LMFA, CNRS, Ecole Centrale Lyon, 69134 Ecully Cedex (France); Gibert, M, E-mail: mathieu.creyssels@ec-lyon.fr [Also at MPI-DS (LFPN) Gottingen (Germany)

    2011-12-22

    To investigate the influence of a rough-wall boundary layer on turbulent heat transport, an experiment of high-Rayleigh convection in water is carried out in a Rayleigh-Benard cell with a rough lower plate and a smooth upper plate. A transition in the heat transport is observed when the thermal boundary layer thickness becomes comparable to or smaller than the roughness height. Besides, at larger Rayleigh numbers than the threshold value, heat transport is found to be increased up to 60%. This enhancement cannot be explained simply by an increase in the contact area of the rough surface since the contact area is increased only by a factor of 40%. Finally, a simple model is proposed to explain the enhanced heat transport.

  3. Surface excitation parameter for rough surfaces

    Da, Bo; Salma, Khanam; Ji, Hui; Mao, Shifeng; Zhang, Guanghui; Wang, Xiaoping; Ding, Zejun

    2015-01-01

    Graphical abstract: - Highlights: • Instead of providing a general mathematical model of roughness, we directly use a finite element triangle mesh method to build a fully 3D rough surface from the practical sample. • The surface plasmon excitation can be introduced to the realistic sample surface by dielectric response theory and finite element method. • We found that SEP calculated based on ideal plane surface model are still reliable for real sample surface with common roughness. - Abstract: In order to assess quantitatively the importance of surface excitation effect in surface electron spectroscopy measurement, surface excitation parameter (SEP) has been introduced to describe the surface excitation probability as an average number of surface excitations that electrons can undergo when they move through solid surface either in incoming or outgoing directions. Meanwhile, surface roughness is an inevitable issue in experiments particularly when the sample surface is cleaned with ion beam bombardment. Surface roughness alters not only the electron elastic peak intensity but also the surface excitation intensity. However, almost all of the popular theoretical models for determining SEP are based on ideal plane surface approximation. In order to figure out whether this approximation is efficient or not for SEP calculation and the scope of this assumption, we proposed a new way to determine the SEP for a rough surface by a Monte Carlo simulation of electron scattering process near to a realistic rough surface, which is modeled by a finite element analysis method according to AFM image. The elastic peak intensity is calculated for different electron incident and emission angles. Assuming surface excitations obey the Poisson distribution the SEPs corrected for surface roughness are then obtained by analyzing the elastic peak intensity for several materials and for different incident and emission angles. It is found that the surface roughness only plays an

  4. Small-Scale Surf Zone Geometric Roughness

    2017-12-01

    using stereo imagery techniques. A waterproof two- camera system with self-logging and internal power was developed using commercial-off-the-shelf...estimates. 14. SUBJECT TERMS surface roughness, nearshore, aerodynamic roughness, surf zone, structure from motion, 3D imagery 15. NUMBER OF... power was developed using commercial-off-the- shelf components and commercial software for operations 1m above the sea surface within the surf zone

  5. How supercontinents and superoceans affect seafloor roughness.

    Whittaker, Joanne M; Müller, R Dietmar; Roest, Walter R; Wessel, Paul; Smith, Walter H F

    2008-12-18

    Seafloor roughness varies considerably across the world's ocean basins and is fundamental to controlling the circulation and mixing of heat in the ocean and dissipating eddy kinetic energy. Models derived from analyses of active mid-ocean ridges suggest that ocean floor roughness depends on seafloor spreading rates, with rougher basement forming below a half-spreading rate threshold of 30-35 mm yr(-1) (refs 4, 5), as well as on the local interaction of mid-ocean ridges with mantle plumes or cold-spots. Here we present a global analysis of marine gravity-derived roughness, sediment thickness, seafloor isochrons and palaeo-spreading rates of Cretaceous to Cenozoic ridge flanks. Our analysis reveals that, after eliminating effects related to spreading rate and sediment thickness, residual roughness anomalies of 5-20 mGal remain over large swaths of ocean floor. We found that the roughness as a function of palaeo-spreading directions and isochron orientations indicates that most of the observed excess roughness is not related to spreading obliquity, as this effect is restricted to relatively rare occurrences of very high obliquity angles (>45 degrees ). Cretaceous Atlantic ocean floor, formed over mantle previously overlain by the Pangaea supercontinent, displays anomalously low roughness away from mantle plumes and is independent of spreading rates. We attribute this observation to a sub-Pangaean supercontinental mantle temperature anomaly leading to slightly thicker than normal Late Jurassic and Cretaceous Atlantic crust, reduced brittle fracturing and smoother basement relief. In contrast, ocean crust formed above Pacific superswells, probably reflecting metasomatized lithosphere underlain by mantle at only slightly elevated temperatures, is not associated with basement roughness anomalies. These results highlight a fundamental difference in the nature of large-scale mantle upwellings below supercontinents and superoceans, and their impact on oceanic crustal

  6. Role of surface roughness in superlubricity

    Tartaglino, U; Samoilov, V N; Persson, B N J

    2006-01-01

    We study the sliding of elastic solids in adhesive contact with flat and rough interfaces. We consider the dependence of the sliding friction on the elastic modulus of the solids. For elastically hard solids with planar surfaces with incommensurate surface structures we observe extremely low friction (superlubricity), which very abruptly increases as the elastic modulus decreases. We show that even a relatively small surface roughness may completely kill the superlubricity state

  7. Wall roughness induces asymptotic ultimate turbulence

    Zhu, Xiaojue; Verschoof, Ruben A.; Bakhuis, Dennis; Huisman, Sander G.; Verzicco, Roberto; Sun, Chao; Lohse, Detlef

    2018-04-01

    Turbulence governs the transport of heat, mass and momentum on multiple scales. In real-world applications, wall-bounded turbulence typically involves surfaces that are rough; however, characterizing and understanding the effects of wall roughness on turbulence remains a challenge. Here, by combining extensive experiments and numerical simulations, we examine the paradigmatic Taylor-Couette system, which describes the closed flow between two independently rotating coaxial cylinders. We show how wall roughness greatly enhances the overall transport properties and the corresponding scaling exponents associated with wall-bounded turbulence. We reveal that if only one of the walls is rough, the bulk velocity is slaved to the rough side, due to the much stronger coupling to that wall by the detaching flow structures. If both walls are rough, the viscosity dependence is eliminated, giving rise to asymptotic ultimate turbulence—the upper limit of transport—the existence of which was predicted more than 50 years ago. In this limit, the scaling laws can be extrapolated to arbitrarily large Reynolds numbers.

  8. Dissolution of minerals with rough surfaces

    de Assis, Thiago A.; Aarão Reis, Fábio D. A.

    2018-05-01

    We study dissolution of minerals with initial rough surfaces using kinetic Monte Carlo simulations and a scaling approach. We consider a simple cubic lattice structure, a thermally activated rate of detachment of a molecule (site), and rough surface configurations produced by fractional Brownian motion algorithm. First we revisit the problem of dissolution of initial flat surfaces, in which the dissolution rate rF reaches an approximately constant value at short times and is controlled by detachment of step edge sites. For initial rough surfaces, the dissolution rate r at short times is much larger than rF ; after dissolution of some hundreds of molecular layers, r decreases by some orders of magnitude across several time decades. Meanwhile, the surface evolves through configurations of decreasing energy, beginning with dissolution of isolated sites, then formation of terraces with disordered boundaries, their growth, and final smoothing. A crossover time to a smooth configuration is defined when r = 1.5rF ; the surface retreat at the crossover is approximately 3 times the initial roughness and is temperature-independent, while the crossover time is proportional to the initial roughness and is controlled by step-edge site detachment. The initial dissolution process is described by the so-called rough rates, which are measured for fixed ratios between the surface retreat and the initial roughness. The temperature dependence of the rough rates indicates control by kink site detachment; in general, it suggests that rough rates are controlled by the weakest microscopic bonds during the nucleation and formation of the lowest energy configurations of the crystalline surface. Our results are related to recent laboratory studies which show enhanced dissolution in polished calcite surfaces. In the application to calcite dissolution in alkaline environment, the minimal values of recently measured dissolution rate spectra give rF ∼10-9 mol/(m2 s), and the calculated rate

  9. Rough mill simulator version 3.0: an analysis tool for refining rough mill operations

    Edward Thomas; Joel Weiss

    2006-01-01

    ROMI-3 is a rough mill computer simulation package designed to be used by both rip-first and chop-first rough mill operators and researchers. ROMI-3 allows users to model and examine the complex relationships among cutting bill, lumber grade mix, processing options, and their impact on rough mill yield and efficiency. Integrated into the ROMI-3 software is a new least-...

  10. Mitigating mask roughness via pupil filtering

    Baylav, B.; Maloney, C.; Levinson, Z.; Bekaert, J.; Vaglio Pret, A.; Smith, B.

    2014-03-01

    The roughness present on the sidewalls of lithographically defined patterns imposes a very important challenge for advanced technology nodes. It can originate from the aerial image or the photoresist chemistry/processing [1]. The latter remains to be the dominant group in ArF and KrF lithography; however, the roughness originating from the mask transferred to the aerial image is gaining more attention [2-9], especially for the imaging conditions with large mask error enhancement factor (MEEF) values. The mask roughness contribution is usually in the low frequency range, which is particularly detrimental to the device performance by causing variations in electrical device parameters on the same chip [10-12]. This paper explains characteristic differences between pupil plane filtering in amplitude and in phase for the purpose of mitigating mask roughness transfer under interference-like lithography imaging conditions, where onedirectional periodic features are to be printed by partially coherent sources. A white noise edge roughness was used to perturbate the mask features for validating the mitigation.

  11. Development of nano-roughness calibration standards

    Baršić, Gorana; Mahović, Sanjin; Zorc, Hrvoje

    2012-01-01

    At the Laboratory for Precise Measurements of Length, currently the Croatian National Laboratory for Length, unique nano-roughness calibration standards were developed, which have been physically implemented in cooperation with the company MikroMasch Trading OU and the Ruđer Bošković Institute. In this paper, a new design for a calibration standard with two measuring surfaces is presented. One of the surfaces is for the reproduction of roughness parameters, while the other is for the traceability of length units below 50 nm. The nominal values of the groove depths on these measuring surfaces are the same. Thus, a link between the measuring surfaces has been ensured, which makes these standards unique. Furthermore, the calibration standards available on the market are generally designed specifically for individual groups of measuring instrumentation, such as interferometric microscopes, stylus instruments, scanning electron microscopes (SEM) or scanning probe microscopes. In this paper, a new design for nano-roughness standards has been proposed for use in the calibration of optical instruments, as well as for stylus instruments, SEM, atomic force microscopes and scanning tunneling microscopes. Therefore, the development of these new nano-roughness calibration standards greatly contributes to the reproducibility of the results of groove depth measurement as well as the 2D and 3D roughness parameters obtained by various measuring methods. (paper)

  12. A Rough Set Approach for Customer Segmentation

    Prabha Dhandayudam

    2014-04-01

    Full Text Available Customer segmentation is a process that divides a business's total customers into groups according to their diversity of purchasing behavior and characteristics. The data mining clustering technique can be used to accomplish this customer segmentation. This technique clusters the customers in such a way that the customers in one group behave similarly when compared to the customers in other groups. The customer related data are categorical in nature. However, the clustering algorithms for categorical data are few and are unable to handle uncertainty. Rough set theory (RST is a mathematical approach that handles uncertainty and is capable of discovering knowledge from a database. This paper proposes a new clustering technique called MADO (Minimum Average Dissimilarity between Objects for categorical data based on elements of RST. The proposed algorithm is compared with other RST based clustering algorithms, such as MMR (Min-Min Roughness, MMeR (Min Mean Roughness, SDR (Standard Deviation Roughness, SSDR (Standard deviation of Standard Deviation Roughness, and MADE (Maximal Attributes DEpendency. The results show that for the real customer data considered, the MADO algorithm achieves clusters with higher cohesion, lower coupling, and less computational complexity when compared to the above mentioned algorithms. The proposed algorithm has also been tested on a synthetic data set to prove that it is also suitable for high dimensional data.

  13. Turbulent flow velocity distribution at rough walls

    Baumann, W.

    1978-08-01

    Following extensive measurements of the velocity profile in a plate channel with artificial roughness geometries specific investigations were carried out to verify the results obtained. The wall geometry used was formed by high transverse square ribs having a large pitch. The measuring position relative to the ribs was varied as a parameter thus providing a statement on the local influence of roughness ribs on the values measured. As a fundamental result it was found that the gradient of the logarithmic rough wall velocity profiles, which differs widely from the value 2.5, depends but slightly on the measuring position relative to the ribs. The gradients of the smooth wall velocity profiles deviate from 2.5 near the ribs, only. This fact can be explained by the smooth wall shear stress varying with the pitch of the ribs. (orig.) 891 GL [de

  14. Spin Hall effect by surface roughness

    Zhou, Lingjun

    2015-01-08

    The spin Hall and its inverse effects, driven by the spin orbit interaction, provide an interconversion mechanism between spin and charge currents. Since the spin Hall effect generates and manipulates spin current electrically, to achieve a large effect is becoming an important topic in both academia and industries. So far, materials with heavy elements carrying a strong spin orbit interaction, provide the only option. We propose here a new mechanism, using the surface roughness in ultrathin films, to enhance the spin Hall effect without heavy elements. Our analysis based on Cu and Al thin films suggests that surface roughness is capable of driving a spin Hall angle that is comparable to that in bulk Au. We also demonstrate that the spin Hall effect induced by surface roughness subscribes only to the side-jump contribution but not the skew scattering. The paradigm proposed in this paper provides the second, not if only, alternative to generate a sizable spin Hall effect.

  15. Why do rough surfaces appear glossy?

    Qi, Lin; Chantler, Mike J; Siebert, J Paul; Dong, Junyu

    2014-05-01

    The majority of work on the perception of gloss has been performed using smooth surfaces (e.g., spheres). Previous studies that have employed more complex surfaces reported that increasing mesoscale roughness increases perceived gloss [Psychol. Sci.19, 196 (2008), J. Vis.10(9), 13 (2010), Curr. Biol.22, 1909 (2012)]. We show that the use of realistic rendering conditions is important and that, in contrast to [Psychol. Sci.19, 196 (2008), J. Vis.10(9), 13 (2010)], after a certain point increasing roughness further actually reduces glossiness. We investigate five image statistics of estimated highlights and show that for our stimuli, one in particular, which we term "percentage of highlight area," is highly correlated with perceived gloss. We investigate a simple model that explains the unimodal, nonmonotonic relationship between mesoscale roughness and percentage highlight area.

  16. Rough estimate demand of atomic energy-related budget for fiscal year 1996

    Kitagishi, Tatsuro

    1996-01-01

    The rough estimate demand of the budget for fiscal year 1996 of eight atomic energy-related ministries and agencies was determined at about 494,879 million yen, which is 2.4% growth as compared with that for the previous year. Concretely, the general account is 204,594 million yen, 2.2% growth, and the special account is 290,285 million yen, 2.6% growth. The budget is 357,060 million yen and 3.7% growth for Science and Technology Agency, 130, 787 million yen and 2% decrease for Ministry of International Trade and Industry, and 7,032 million yen and 29.2% increase for other six ministries and agencies. Emphasis is placed on the research of upgrading LWRs including the disassembling of reactors, the performance test for fuel, the improvement of reactor technology and the verifying test of practical reactor decommissioning facilities, and the research and development of advanced nuclear fuel cycle technology. Also the technical development of waste treatment and disposal including high level radioactive waste is carried out with 40.3 billion yen. Atomic Energy Commission exerts efforts for the development of atomic energy policy for the peaceful utilization, the establishment of coordinative LWR power generation system, the development of nuclear fuel recycling and the strengthening of the basic research on atomic energy. (K.I.)

  17. Fuzzy sets, rough sets, multisets and clustering

    Dahlbom, Anders; Narukawa, Yasuo

    2017-01-01

    This book is dedicated to Prof. Sadaaki Miyamoto and presents cutting-edge papers in some of the areas in which he contributed. Bringing together contributions by leading researchers in the field, it concretely addresses clustering, multisets, rough sets and fuzzy sets, as well as their applications in areas such as decision-making. The book is divided in four parts, the first of which focuses on clustering and classification. The second part puts the spotlight on multisets, bags, fuzzy bags and other fuzzy extensions, while the third deals with rough sets. Rounding out the coverage, the last part explores fuzzy sets and decision-making.

  18. Diffuse neutron scattering signatures of rough films

    Pynn, R.; Lujan, M. Jr.

    1992-01-01

    Patterns of diffuse neutron scattering from thin films are calculated from a perturbation expansion based on the distorted-wave Born approximation. Diffuse fringes can be categorised into three types: those that occur at constant values of the incident or scattered neutron wavevectors, and those for which the neutron wavevector transfer perpendicular to the film is constant. The variation of intensity along these fringes can be used to deduce the spectrum of surface roughness for the film and the degree of correlation between the film's rough surfaces

  19. Total sensitivity and uncertainty analysis for LWR pin-cells with improved UNICORN code

    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.

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

    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

  1. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

    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. Evaluating the loss of a LWR spent fuel or plutonium shipping package into the sea

    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,

  3. Safety-related investigations on power distribution in MOX fuel elements in LWR cores

    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

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

    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.

  5. Residual life assessment of major LWR components: NPAR approach and results

    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)

  6. Prioritization of tasks in the draft LWR safety technology program plan. Final report

    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

  7. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    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

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

    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

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

    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

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

    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

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

    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.)

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

    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

  13. Advantages of retrofitting high velocity separators to LWR turbines; experience in VVR NPP Loviisa

    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

  14. Uncertainties in criticality analysis which affect the storage and transportation of LWR fuel

    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

  15. Short Communication on "In-situ TEM ion irradiation investigations on U3Si2 at LWR temperatures"

    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

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

    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

  18. Potential roughness near lithographically fabricated atom chips

    Krüger, Peter; Andersson, L. M.; Wildermuth, Stefan

    2007-01-01

    Potential roughness has been reported to severely impair experiments in magnetic microtraps. We show that these obstacles can be overcome as we measure disorder potentials that are reduced by two orders of magnitude near lithographically patterned high-quality gold layers on semiconductor atom chip...

  19. Rough flows and homogenization in stochastic turbulence

    Bailleul, I.; Catellier, R.

    2016-01-01

    We provide in this work a tool-kit for the study of homogenisation of random ordinary differential equations, under the form of a friendly-user black box based on the tehcnology of rough flows. We illustrate the use of this setting on the example of stochastic turbulence.

  20. Reproducibility of surface roughness in reaming

    Müller, Pavel; De Chiffre, Leonardo

    An investigation on the reproducibility of surface roughness in reaming was performed to document the applicability of this approach for testing cutting fluids. Austenitic stainless steel was used as a workpiece material and HSS reamers as cutting tools. Reproducibility of the results was evaluat...

  1. Optical measurement of surface roughness in manufacturing

    Brodmann, R.

    1984-11-01

    The measuring system described here is based on the light-scattering method, and was developed by Optische Werke G. Rodenstock, Munich. It is especially useful for rapid non-contact monitoring of surface roughness in production-related areas. This paper outlines the differences between this system and the common stylus instrument, including descriptions of some applications in industry.

  2. Microscopic Holography for flow over rough plate

    Talapatra, Siddharth; Hong, Jiarong; Lu, Yuan; Katz, Joseph

    2008-11-01

    Our objective is to measure the near wall flow structures in a turbulent channel flow over a rough wall. In-line microscopic holographic PIV can resolve the 3-D flow field in a small sample volume, but recording holograms through a rough surface is a challenge. To solve this problem, we match the refractive indices of the fluid with that of the wall. Proof of concept tests involve an acrylic plate containing uniformly distributed, closely packed 0.45mm high pyramids with slope angle of 22^^o located within a concentrated sodium iodide solution. Holograms recorded by a 4864 x 3248 pixel digital camera at 10X magnification provide a field of view of 3.47mm x 2.32mm and pixel resolution of 0.714 μm. Due to index matching, reconstructed seed particles can be clearly seen over the entire volume, with only faint traces with the rough wall that can be removed. Planned experiments will be performed in a 20 x 5 cm rectangular channel with the top and bottom plates having the same roughness as the sample plate.

  3. Factors influencing surface roughness of polyimide film

    Yao Hong; Zhang Zhanwen; Huang Yong; Li Bo; Li Sai

    2011-01-01

    The polyimide (PI) films of pyromellitic dianhydride-oxydiamiline (PMDA-ODA) were fabricated using vapor deposition polymerization (VDP) method under high vacuum pressure of 10-4 Pa level. The influence of equipment, substrate temperature, the process of heating and deposition ratio of monomers on the surface roughness of the PI films was investigated. The surface topography of films was measured by interferometer microscopy and scanning electron microscopy(SEM), and the surface roughness was probed with atomic force microscopy(AFM). The results show that consecutive films can be formed when the distance from steering flow pipe to substrate is 74 cm. The surface roughnesses are 291.2 nm and 61.9 nm respectively for one-step heating process and multi-step heating process, and using fine mesh can effectively avoid the splash of materials. The surface roughness can be 3.3 nm when the deposition rate ratio of PMDA to ODA is 0.9:1, and keeping the temperature of substrate around 30 degree C is advantageous to form a film with planar micro-surface topography. (authors)

  4. Roughly isometric minimal immersions into Riemannian manifolds

    Markvorsen, Steen

    of the intrinsic combinatorial discrete Laplacian, and we will show that they share several analytic and geometric properties with their smooth (minimal submanifold) counterparts in $N$. The intrinsic properties thus obtained may hence serve as roughly invariant descriptors for the original metric space $X$....

  5. Three-tier rough superhydrophobic surfaces

    Cao, Yuanzhi; Yuan, Longyan; Hu, Bin; Zhou, Jun

    2015-01-01

    A three-tier rough superhydrophobic surface was fabricated by growing hydrophobic modified (fluorinated silane) zinc oxide (ZnO)/copper oxide (CuO) hetero-hierarchical structures on silicon (Si) micro-pillar arrays. Compared with the other three control samples with a less rough tier, the three-tier surface exhibits the best water repellency with the largest contact angle 161° and the lowest sliding angle 0.5°. It also shows a robust Cassie state which enables the water to flow with a speed over 2 m s"−"1. In addition, it could prevent itself from being wetted by the droplet with low surface tension (mixed water and ethanol 1:1 in volume) which reveals a flow speed of 0.6 m s"−"1 (dropped from the height of 2 cm). All these features prove that adding another rough tier on a two-tier rough surface could futher improve its water-repellent properties. (paper)

  6. Roughness-induced streaming in turbulent wave boundary layers

    Fuhrman, David R.; Sumer, B. Mutlu; Fredsøe, Jørgen

    2011-01-01

    -averaged streaming characteristics induced by bottom roughness variations are systematically assessed. The effects of variable roughness ratio, gradual roughness transitions, as well as changing flow orientation in plan are all considered. As part of the latter, roughness-induced secondary flows are predicted...

  7. Self-affine roughness influence on redox reaction charge admittance

    Palasantzas, G

    2005-01-01

    In this work we investigate the influence of self-affine electrode roughness on the admittance of redox reactions during facile charge transfer kinetics. The self-affine roughness is characterized by the rms roughness amplitude w, the correlation length xi and the roughness exponent H (0

  8. More on neutrosophic soft rough sets and its modification

    Emad Marei

    2015-12-01

    Full Text Available This paper aims to introduce and discuss anew mathematical tool for dealing with uncertainties, which is a combination of neutrosophic sets, soft sets and rough sets, namely neutrosophic soft rough set model. Also, its modification is introduced. Some of their properties are studied and supported with proved propositions and many counter examples. Some of rough relations are redefined as a neutrosophic soft rough relations. Comparisons among traditional rough model, suggested neutrosophic soft rough model and its modification, by using their properties and accuracy measures are introduced. Finally, we illustrate that, classical rough set model can be viewed as a special case of suggested models in this paper.

  9. Turbulent boundary layer over roughness transition with variation in spanwise roughness length scale

    Westerweel, Jerry; Tomas, Jasper; Eisma, Jerke; Pourquie, Mathieu; Elsinga, Gerrit; Jonker, Harm

    2016-11-01

    Both large-eddy simulations (LES) and water-tunnel experiments, using simultaneous stereoscopic PIV and LIF were done to investigate pollutant dispersion in a region where the surface changes from rural to urban roughness. This consists of rectangular obstacles where we vary the spanwise aspect ratio of the obstacles. A line source of passive tracer was placed upstream of the roughness transition. The objectives of the study are: (i) to determine the influence of the aspect ratio on the roughness-transition flow, and (ii) to determine the dominant mechanisms of pollutant removal from street canyons in the transition region. It is found that for a spanwise aspect ratio of 2 the drag induced by the roughness is largest of all considered cases, which is caused by a large-scale secondary flow. In the roughness transition the vertical advective pollutant flux is the main ventilation mechanism in the first three streets. Furthermore, by means of linear stochastic estimation the mean flow structure is identied that is responsible for exchange of the fluid between the roughness obstacles and the outer part of the boundary layer. Furthermore, it is found that the vertical length scale of this structure increases with increasing aspect ratio of the obstacles in the roughness region.

  10. Numerical Investigation of Effect of Surface Roughness in a Microchannel

    Shin, Myung Seob; Byun, Sung Jun; Yoon, Joon Yong [Hanyang University, Seoul (Korea, Republic of)

    2010-05-15

    In this paper, lattice Boltzmann method(LBM) results for a laminar flow in a microchannel with rough surface are presented. The surface roughness is modeled as an array of rectangular modules placed on the top and bottom surface of a parallel-plate channel. The effects of relative surface roughness, roughness distribution, and roughness size are presented in terms of the Poiseuille number. The roughness distribution characterized by the ratio of the roughness height to the spacing between the modules has a negligible effect on the flow and friction factors. Finally, a significant increase in the Poiseuille number is observed when the surface roughness is considered, and the effects of roughness on the microflow field mainly depend on the surface roughness.

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

    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)

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

    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

  13. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    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

  14. Prediction of fission product and aerosol behaviour during a postulated severe accident in a LWR

    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

  15. In-core materials testing under LWR conditions in the Halden reactor

    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)

  16. Data needs for long-term dry storage of LWR fuel. Interim report

    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

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

    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)

  18. Decay heat and gamma dose-rate prediction capability in spent LWR fuel

    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

  19. On the Diffusion Coefficient of Two-step Method for LWR analysis

    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

  20. Long-term embrittlement of cast duplex stainless steels in LWR systems

    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

  1. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    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.)

  2. In-core materials testing under LWR conditions in the Halden reactor

    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)

  3. The Width of High Burnup Structure in LWR UO2 Fuel

    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)

  4. Anomalous roughness of turbulent interfaces with system size dependent local roughness exponent

    Balankin, Alexander S.; Matamoros, Daniel Morales

    2005-01-01

    In a system far from equilibrium the system size can play the role of control parameter that governs the spatiotemporal dynamics of the system. Accordingly, the kinetic roughness of interfaces in systems far from equilibrium may depend on the system size. To get an insight into this problem, we performed a detailed study of rough interfaces formed in paper combustion experiments. Using paper sheets of different width λ, we found that the turbulent flame fronts display anomalous multi-scaling characterized by non-universal global roughness exponent α and by the system size dependent spectrum of local roughness exponents, ζ q (λ)=ζ 1 (1)q -ω λ φ q =0.93q -0.15 . The structure factor of turbulent flame fronts also exhibits unconventional scaling dependence on λ. These results are expected to apply to a broad range of far from equilibrium systems when the kinetic energy fluctuations exceed a certain critical value.

  5. Generalized rough sets hybrid structure and applications

    Mukherjee, Anjan

    2015-01-01

    The book introduces the concept of “generalized interval valued intuitionistic fuzzy soft sets”. It presents the basic properties of these sets and also, investigates an application of generalized interval valued intuitionistic fuzzy soft sets in decision making with respect to interval of degree of preference. The concept of “interval valued intuitionistic fuzzy soft rough sets” is discussed and interval valued intuitionistic fuzzy soft rough set based multi criteria group decision making scheme is presented, which refines the primary evaluation of the whole expert group and enables us to select the optimal object in a most reliable manner. The book also details concept of interval valued intuitionistic fuzzy sets of type 2. It presents the basic properties of these sets. The book also introduces the concept of “interval valued intuitionistic fuzzy soft topological space (IVIFS topological space)” together with intuitionistic fuzzy soft open sets (IVIFS open sets) and intuitionistic fuzzy soft cl...

  6. Single-layer model for surface roughness.

    Carniglia, C K; Jensen, D G

    2002-06-01

    Random roughness of an optical surface reduces its specular reflectance and transmittance by the scattering of light. The reduction in reflectance can be modeled by a homogeneous layer on the surface if the refractive index of the layer is intermediate to the indices of the media on either side of the surface. Such a layer predicts an increase in the transmittance of the surface and therefore does not provide a valid model for the effects of scatter on the transmittance. Adding a small amount of absorption to the layer provides a model that predicts a reduction in both reflectance and transmittance. The absorbing layer model agrees with the predictions of a scalar scattering theory for a layer with a thickness that is twice the rms roughness of the surface. The extinction coefficient k for the layer is proportional to the thickness of the layer.

  7. Offshore Wind Power at Rough Sea

    Petersen, Kristian Rasmus; Madsen, Erik Skov; Bilberg, Arne

    2013-01-01

    This study compare the current operations and maintenance issues of one offshore wind park at very rough sea conditions and two onshore wind parks. Through a detailed data analysis and case studies this study identifies how improvements have been made in maintenance of large wind turbines. Howeve......, the study has also revealed the need for new maintenance models including a shift from breakdown and preventive maintenances and towards more predictive maintenance to reduce the cost of energy for offshore wind energy installations in the future.......This study compare the current operations and maintenance issues of one offshore wind park at very rough sea conditions and two onshore wind parks. Through a detailed data analysis and case studies this study identifies how improvements have been made in maintenance of large wind turbines. However...

  8. The contact sport of rough surfaces

    Carpick, Robert W.

    2018-01-01

    Describing the way two surfaces touch and make contact may seem simple, but it is not. Fully describing the elastic deformation of ideally smooth contacting bodies, under even low applied pressure, involves second-order partial differential equations and fourth-rank elastic constant tensors. For more realistic rough surfaces, the problem becomes a multiscale exercise in surface-height statistics, even before including complex phenomena such as adhesion, plasticity, and fracture. A recent research competition, the “Contact Mechanics Challenge” (1), was designed to test various approximate methods for solving this problem. A hypothetical rough surface was generated, and the community was invited to model contact with this surface with competing theories for the calculation of properties, including contact area and pressure. A supercomputer-generated numerical solution was kept secret until competition entries were received. The comparison of results (2) provides insights into the relative merits of competing models and even experimental approaches to the problem.

  9. Prediction of Ductile Fracture Surface Roughness Scaling

    Needleman, Alan; Tvergaard, Viggo; Bouchaud, Elisabeth

    2012-01-01

    . Ductile crack growth in a thin strip under mode I, overall plane strain, small scale yielding conditions is analyzed. Although overall plane strain loading conditions are prescribed, full 3D analyses are carried out to permit modeling of the three dimensional material microstructure and of the resulting......Experimental observations have shown that the roughness of fracture surfaces exhibit certain characteristic scaling properties. Here, calculations are carried out to explore the extent to which a ductile damage/fracture constitutive relation can be used to model fracture surface roughness scaling...... three dimensional stress and deformation states that develop in the fracture process region. An elastic-viscoplastic constitutive relation for a progressively cavitating plastic solid is used to model the material. Two populations of second phase particles are represented: large inclusions with low...

  10. Estimation of gloss from rough surface parameters

    Simonsen, Ingve; Larsen, Åge G.; Andreassen, Erik; Ommundsen, Espen; Nord-Varhaug, Katrin

    2005-12-01

    Gloss is a quantity used in the optical industry to quantify and categorize materials according to how well they scatter light specularly. With the aid of phase perturbation theory, we derive an approximate expression for this quantity for a one-dimensional randomly rough surface. It is demonstrated that gloss depends in an exponential way on two dimensionless quantities that are associated with the surface randomness: the root-mean-square roughness times the perpendicular momentum transfer for the specular direction, and a correlation function dependent factor times a lateral momentum variable associated with the collection angle. Rigorous Monte Carlo simulations are used to access the quality of this approximation, and good agreement is observed over large regions of parameter space.

  11. Sparseness and Roughness of Foreign Exchange Rates

    Vandewalle, N.; Ausloos, M.

    An accurate multiaffine analysis of 23 foreign currency exchange rates has been performed. The roughness exponent H1 which characterizes the excursion of the exchange rate has been numerically measured. The degree of intermittency C1 has been also estimated. In the (H1,C1) phase diagram, the currency exchange rates are dispersed in a wide region around the Brownian motion value (H1=0.5,C1=0) and have a significantly intermittent component (C1≠0).

  12. Rough surface scattering simulations using graphics cards

    Klapetek, Petr; Valtr, Miroslav; Poruba, Ales; Necas, David; Ohlidal, Miloslav

    2010-01-01

    In this article we present results of rough surface scattering calculations using a graphical processing unit implementation of the Finite Difference in Time Domain algorithm. Numerical results are compared to real measurements and computational performance is compared to computer processor implementation of the same algorithm. As a basis for computations, atomic force microscope measurements of surface morphology are used. It is shown that the graphical processing unit capabilities can be used to speedup presented computationally demanding algorithms without loss of precision.

  13. Roughness Length Variability over Heterogeneous Surfaces

    2010-03-01

    2004), the influence of variable roughness reaches its maximum at the height of local 0z and vanishes at the so- called blending height (Wieringa...the distribution of visibility restrictors such as low clouds, fog, haze, dust, and pollutants . An improved understanding of ABL structure...R. D., B. H. Lynn, A. Boone, W.-K. Tao, and J. Simpson, 2001: The influence of soil moisture, coastline curvature, and land-breeze circulations on

  14. Analysis of accuracy in photogrammetric roughness measurements

    Olkowicz, Marcin; Dąbrowski, Marcin; Pluymakers, Anne

    2017-04-01

    Regarding permeability, one of the most important features of shale gas reservoirs is the effective aperture of cracks opened during hydraulic fracturing, both propped and unpropped. In a propped fracture, the aperture is controlled mostly by proppant size and its embedment, and fracture surface roughness only has a minor influence. In contrast, in an unpropped fracture aperture is controlled by the fracture roughness and the wall displacement. To measure fracture surface roughness, we have used the photogrammetric method since it is time- and cost-efficient. To estimate the accuracy of this method we compare the photogrammetric measurements with reference measurements taken with a White Light Interferometer (WLI). Our photogrammetric setup is based on high resolution 50 Mpx camera combined with a focus stacking technique. The first step for photogrammetric measurements is to determine the optimal camera positions and lighting. We compare multiple scans of one sample, taken with different settings of lighting and camera positions, with the reference WLI measurement. The second step is to perform measurements of all studied fractures with the parameters that produced the best results in the first step. To compare photogrammetric and WLI measurements we regrid both data sets onto a regular 10 μm grid and determined the best fit, followed by a calculation of the difference between the measurements. The first results of the comparison show that for 90 % of measured points the absolute vertical distance between WLI and photogrammetry is less than 10 μm, while the mean absolute vertical distance is 5 μm. This proves that our setup can be used for fracture roughness measurements in shales.

  15. The characteristic function of rough Heston models

    Euch, Omar El; Rosenbaum, Mathieu

    2016-01-01

    It has been recently shown that rough volatility models, where the volatility is driven by a fractional Brownian motion with small Hurst parameter, provide very relevant dynamics in order to reproduce the behavior of both historical and implied volatilities. However, due to the non-Markovian nature of the fractional Brownian motion, they raise new issues when it comes to derivatives pricing. Using an original link between nearly unstable Hawkes processes and fractional volatility models, we c...

  16. Radiative transfer model for contaminated rough slabs.

    Andrieu, François; Douté, Sylvain; Schmidt, Frédéric; Schmitt, Bernard

    2015-11-01

    We present a semi-analytical model to simulate the bidirectional reflectance distribution function (BRDF) of a rough slab layer containing impurities. This model has been optimized for fast computation in order to analyze massive hyperspectral data by a Bayesian approach. We designed it for planetary surface ice studies but it could be used for other purposes. It estimates the bidirectional reflectance of a rough slab of material containing inclusions, overlaying an optically thick media (semi-infinite media or stratified media, for instance granular material). The inclusions are assumed to be close to spherical and constituted of any type of material other than the ice matrix. It can be any other type of ice, mineral, or even bubbles defined by their optical constants. We assume a low roughness and we consider the geometrical optics conditions. This model is thus applicable for inclusions larger than the considered wavelength. The scattering on the inclusions is assumed to be isotropic. This model has a fast computation implementation and thus is suitable for high-resolution hyperspectral data analysis.

  17. Multi-decadal Arctic sea ice roughness.

    Tsamados, M.; Stroeve, J.; Kharbouche, S.; Muller, J. P., , Prof; Nolin, A. W.; Petty, A.; Haas, C.; Girard-Ardhuin, F.; Landy, J.

    2017-12-01

    The transformation of Arctic sea ice from mainly perennial, multi-year ice to a seasonal, first-year ice is believed to have been accompanied by a reduction of the roughness of the ice cover surface. This smoothening effect has been shown to (i) modify the momentum and heat transfer between the atmosphere and ocean, (ii) to alter the ice thickness distribution which in turn controls the snow and melt pond repartition over the ice cover, and (iii) to bias airborne and satellite remote sensing measurements that depend on the scattering and reflective characteristics over the sea ice surface topography. We will review existing and novel remote sensing methodologies proposed to estimate sea ice roughness, ranging from airborne LIDAR measurement (ie Operation IceBridge), to backscatter coefficients from scatterometers (ASCAT, QUICKSCAT), to multi angle maging spectroradiometer (MISR), and to laser (Icesat) and radar altimeters (Envisat, Cryosat, Altika, Sentinel-3). We will show that by comparing and cross-calibrating these different products we can offer a consistent multi-mission, multi-decadal view of the declining sea ice roughness. Implications for sea ice physics, climate and remote sensing will also be discussed.

  18. ROUGHNESS ANALYSIS OF VARIOUSLY POLISHED NIOBIUM SURFACES

    Ribeill, G.; Reece, C.

    2008-01-01

    Niobium superconducting radio frequency (SRF) cavities have gained widespread use in accelerator systems. It has been shown that surface roughness is a determining factor in the cavities’ effi ciency and maximum accelerating potential achievable through this technology. Irregularities in the surface can lead to spot heating, undesirable local electrical fi eld enhancement and electron multipacting. Surface quality is typically ensured through the use of acid etching in a Buffered Chemical Polish (BCP) bath and electropolishing (EP). In this study, the effects of these techniques on surface morphology have been investigated in depth. The surface of niobium samples polished using different combinations of these techniques has been characterized through atomic force microscopy (AFM) and stylus profi lometry across a range of length scales. The surface morphology was analyzed using spectral techniques to determine roughness and characteristic dimensions. Experimentation has shown that this method is a valuable tool that provides quantitative information about surface roughness at different length scales. It has demonstrated that light BCP pretreatment and lower electrolyte temperature favors a smoother electropolish. These results will allow for the design of a superior polishing process for niobium SRF cavities and therefore increased accelerator operating effi ciency and power.

  19. Modeling superhydrophobic surfaces comprised of random roughness

    Samaha, M. A.; Tafreshi, H. Vahedi; Gad-El-Hak, M.

    2011-11-01

    We model the performance of superhydrophobic surfaces comprised of randomly distributed roughness that resembles natural surfaces, or those produced via random deposition of hydrophobic particles. Such a fabrication method is far less expensive than ordered-microstructured fabrication. The present numerical simulations are aimed at improving our understanding of the drag reduction effect and the stability of the air-water interface in terms of the microstructure parameters. For comparison and validation, we have also simulated the flow over superhydrophobic surfaces made up of aligned or staggered microposts for channel flows as well as streamwise or spanwise ridge configurations for pipe flows. The present results are compared with other theoretical and experimental studies. The numerical simulations indicate that the random distribution of surface roughness has a favorable effect on drag reduction, as long as the gas fraction is kept the same. The stability of the meniscus, however, is strongly influenced by the average spacing between the roughness peaks, which needs to be carefully examined before a surface can be recommended for fabrication. Financial support from DARPA, contract number W91CRB-10-1-0003, is acknowledged.

  20. Rough – Granular Computing knowledge discovery models

    Mohammed M. Eissa

    2016-11-01

    Full Text Available Medical domain has become one of the most important areas of research in order to richness huge amounts of medical information about the symptoms of diseases and how to distinguish between them to diagnose it correctly. Knowledge discovery models play vital role in refinement and mining of medical indicators to help medical experts to settle treatment decisions. This paper introduces four hybrid Rough – Granular Computing knowledge discovery models based on Rough Sets Theory, Artificial Neural Networks, Genetic Algorithm and Rough Mereology Theory. A comparative analysis of various knowledge discovery models that use different knowledge discovery techniques for data pre-processing, reduction, and data mining supports medical experts to extract the main medical indicators, to reduce the misdiagnosis rates and to improve decision-making for medical diagnosis and treatment. The proposed models utilized two medical datasets: Coronary Heart Disease dataset and Hepatitis C Virus dataset. The main purpose of this paper was to explore and evaluate the proposed models based on Granular Computing methodology for knowledge extraction according to different evaluation criteria for classification of medical datasets. Another purpose is to make enhancement in the frame of KDD processes for supervised learning using Granular Computing methodology.

  1. 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

    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)

  2. The surface roughness and planetary boundary layer

    Telford, James W.

    1980-03-01

    Applications of the entrainment process to layers at the boundary, which meet the self similarity requirements of the logarithmic profile, have been studied. By accepting that turbulence has dominating scales related in scale length to the height above the surface, a layer structure is postulated wherein exchange is rapid enough to keep the layers internally uniform. The diffusion rate is then controlled by entrainment between layers. It has been shown that theoretical relationships derived on the basis of using a single layer of this type give quantitatively correct factors relating the turbulence, wind and shear stress for very rough surface conditions. For less rough surfaces, the surface boundary layer can be divided into several layers interacting by entrainment across each interface. This analysis leads to the following quantitatively correct formula compared to published measurements. 1 24_2004_Article_BF00877766_TeX2GIFE1.gif {σ _w }/{u^* } = ( {2/{9Aa}} )^{{1/4}} ( {1 - 3^{{1/2}{ a/k{d_n }/z{σ _w }/{u^* }z/L} )^{{1/4}} = 1.28(1 - 0.945({{σ _w }/{u^* }}}) {{z/L}})^{{1/4 where u^* = ( {{tau/ρ}}^{{1/2}}, σ w is the standard deviation of the vertical velocity, z is the height and L is the Obukhov scale lenght. The constants a, A, k and d n are the entrainment constant, the turbulence decay constant, Von Karman's constant, and the layer depth derived from the theory. Of these, a and A, are universal constants and not empirically determined for the boundary layer. Thus the turbulence needed for the plume model of convection, which resides above these layers and reaches to the inversion, is determined by the shear stress and the heat flux in the surface layers. This model applies to convection in cool air over a warm sea. The whole field is now determined except for the temperature of the air relative to the water, and the wind, which need a further parameter describing sea surface roughness. As a first stop to describing a surface where roughness elements

  3. Modelling intragranular fission gas release in irradiation of sintered LWR UO2 fuel

    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

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

    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

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

    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.

  6. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    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

  7. Verification of the depletion capabilities of the MCNPX code on a LWR MOX fuel assembly

    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)

  8. Overview of LWR severe accident research activities at the Karlsruhe Institute of Technology

    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)

  9. Assessment of management alternatives for LWR wastes. Volume 1. Main achievements of the joint study

    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)

  10. Cracking in LWR RPV head penetrations. Working material. Proceedings of a specialists meeting

    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.

  11. Critical corrosion issues and mitigation strategies impacting the operability of LWR's

    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

  12. Evaluation of alternative waste management schemes for LWR hulls and caps

    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

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

    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.

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

    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)

  15. Improving the computation efficiency of COBRA-TF for LWR safety analysis of large problems

    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)

  16. Descriptions of reference LWR facilities for analysis of nuclear fuel cycles

    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.

  17. Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study

    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.

  18. Recent results from CEC cost sharing research programme on LWR fuel behaviour under accident conditions

    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)

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

    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.

  20. Cracking in LWR RPV head penetrations. Working material. Proceedings of a specialists meeting

    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

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

    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.

  2. Mechanical damage due to corrosion of parts of pump technology and valves of LWR power installations

    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.)

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

    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

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

    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

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

    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

  6. Vectorization of LWR transient analysis code RELAP5/MOD1 and its effect

    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)

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

    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

  8. Plant for retention of 14C in reprocessing plants for LWR fuel elements

    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

  9. Aging assessment and mitigation for major LWR [light water reactor] components

    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

  10. Inner-outer interactions in a rough wall turbulent boundary layer over hemispherical roughness using PIV

    Pathikonda, Gokul; Clark, Caitlyn; Christensen, Kenneth T.

    2017-11-01

    Inner-outer interactions over rough-wall boundary layer were investigated using high frame-rate, PIV measurements in a Refractive index-matched (RIM) facility. Flows over canonical smooth-wall and hexagonally-packed hemispherical roughness under transitionally rough flow conditions (and with Reτ 1500) were measured using a dual camera PIV system with different fields of view (FOVs) and operating simultaneously. The large FOV measures the large scales and boundary layer parameters, while the small FOV measures the small scales very close to the wall with high spatial ( 7y*) and temporal ( 2.5t*) resolutions. Conditional metrics were formulated to investigate these scale interactions in a spatio-temporal sense using the PIV data. It was found that the observations complement the interaction structure made via hotwire experiments and DNS in previous studies over both smooth and rough-wall flows, with a strong correlation between the large scales and small scale energies indicative of the amplitude modulation interactions. Additionally, frequency and scale modulations were also investigated with limited success. These experiments highlight the similarities and differences in these interactions between the smooth- and rough-wall flows.

  11. Dewetting of thin polymer film on rough substrate: II. Experiment

    Volodin, Pylyp; Kondyurin, Alexey

    2008-01-01

    The theory of the dewetting process developed for a model of substrate-film interaction forces was examined by an experimental investigation of the dewetting process of thin polystyrene (PS) films on chemically etched silicon substrates. In the dependence on PS films thickness and silicon roughness, various situations of dewetting were observed as follows: (i) if the wavelength of the substrate roughness is much larger than the critical spinodal wavelength of a film, then spinodal dewetting of the film is observed; (ii) if the wavelength of the substrate roughness is smaller than the critical wavelength of the film and the substrate roughness is larger in comparison with film thickness, then the dewetting due to substrate roughness is observed and the dewetted film patterns repeat the rough substrate structure; (iii) if the wavelength of the substrate roughness is smaller than the critical wavelength of the film and the substrate roughness is small in comparison with the film thickness, then spinodal dewetting proceeds

  12. Sub-Patch Roughness in Earthquake Rupture Investigations

    Zielke, Olaf; Mai, Paul Martin

    2016-01-01

    Fault geometric complexities exhibit fractal characteristics over a wide range of spatial scales (<µm to >km) and strongly affect the rupture process at corresponding scales. Numerical rupture simulations provide a framework to quantitatively investigate the relationship between a fault's roughness and its seismic characteristics. Fault discretization however introduces an artificial lower limit to roughness. Individual fault patches are planar and sub-patch roughnessroughness at spatial scales below fault-patch size– is not incorporated. Does negligence of sub-patch roughness measurably affect the outcome of earthquake rupture simulations? We approach this question with a numerical parameter space investigation and demonstrate that sub-patch roughness significantly modifies the slip-strain relationship –a fundamental aspect of dislocation theory. Faults with sub-patch roughness induce less strain than their planar-fault equivalents at distances beyond the length of a slipping fault. We further provide regression functions that characterize the stochastic effect sub-patch roughness.

  13. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    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.

  14. Assessment of nitrogen as an atmosphere for dry storage of spent LWR fuel

    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

  15. New Developments in Actinides Burning with Symbiotic LWR-HTR-GCFR Fuel Cycles

    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

  16. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    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

  17. Roughness characterization of the galling of metals

    Hubert, C.; Marteau, J.; Deltombe, R.; Chen, Y. M.; Bigerelle, M.

    2014-09-01

    Several kinds of tests exist to characterize the galling of metals, such as that specified in ASTM Standard G98. While the testing procedure is accurate and robust, the analysis of the specimen's surfaces (area=1.2 cm) for the determination of the critical pressure of galling remains subject to operator judgment. Based on the surface's topography analyses, we propose a methodology to express the probability of galling according to the macroscopic pressure load. After performing galling tests on 304L stainless steel, a two-step segmentation of the S q parameter (root mean square of surface amplitude) computed from local roughness maps (100 μ m× 100 μ m) enables us to distinguish two tribological processes. The first step represents the abrasive wear (erosion) and the second one the adhesive wear (galling). The total areas of both regions are highly relevant to quantify galling and erosion processes. Then, a one-parameter phenomenological model is proposed to objectively determine the evolution of non-galled relative area A e versus the pressure load P, with high accuracy ({{A}e}=100/(1+a{{P}2}) with a={{0.54}+/- 0.07}× {{10}-3} M P{{a}-2} and with {{R}2}=0.98). From this model, the critical pressure of galling is found to be equal to 43MPa. The {{S}5 V} roughness parameter (the five deepest valleys in the galled region's surface) is the most relevant roughness parameter for the quantification of damages in the ‘galling region’. The significant valleys’ depths increase from 10 μm-250 μm when the pressure increases from 11-350 MPa, according to a power law ({{S}5 V}=4.2{{P}0.75}, with {{R}2}=0.93).

  18. Robust surface roughness indices and morphological interpretation

    Trevisani, Sebastiano; Rocca, Michele

    2016-04-01

    Geostatistical-based image/surface texture indices based on variogram (Atkison and Lewis, 2000; Herzfeld and Higginson, 1996; Trevisani et al., 2012) and on its robust variant MAD (median absolute differences, Trevisani and Rocca, 2015) offer powerful tools for the analysis and interpretation of surface morphology (potentially not limited to solid earth). In particular, the proposed robust index (Trevisani and Rocca, 2015) with its implementation based on local kernels permits the derivation of a wide set of robust and customizable geomorphometric indices capable to outline specific aspects of surface texture. The stability of MAD in presence of signal noise and abrupt changes in spatial variability is well suited for the analysis of high-resolution digital terrain models. Moreover, the implementation of MAD by means of a pixel-centered perspective based on local kernels, with some analogies to the local binary pattern approach (Lucieer and Stein, 2005; Ojala et al., 2002), permits to create custom roughness indices capable to outline different aspects of surface roughness (Grohmann et al., 2011; Smith, 2015). In the proposed poster, some potentialities of the new indices in the context of geomorphometry and landscape analysis will be presented. At same time, challenges and future developments related to the proposed indices will be outlined. Atkinson, P.M., Lewis, P., 2000. Geostatistical classification for remote sensing: an introduction. Computers & Geosciences 26, 361-371. Grohmann, C.H., Smith, M.J., Riccomini, C., 2011. Multiscale Analysis of Topographic Surface Roughness in the Midland Valley, Scotland. IEEE Transactions on Geoscience and Remote Sensing 49, 1220-1213. Herzfeld, U.C., Higginson, C.A., 1996. Automated geostatistical seafloor classification - Principles, parameters, feature vectors, and discrimination criteria. Computers and Geosciences, 22 (1), pp. 35-52. Lucieer, A., Stein, A., 2005. Texture-based landform segmentation of LiDAR imagery

  19. Traceability of optical roughness measurements on polymers

    De Chiffre, Leonardo; Gasparin, Stefania; Carli, Lorenzo

    2008-01-01

    -focus instrument, and a confocal microscope. Using stylus measurements as reference, parameter settings on the optical instruments were optimised and residual noise reduced by low pass filtering. Traceability of optical measurements could be established with expanded measuring uncertainties (k=2) of 4......An experimental investigation on surface roughness measurements on plastics was carried out with the objective of developing a methodology to achieve traceability of optical instruments. A ground steel surface and its replicas were measured using a stylus instrument, an optical auto......% for the auto-focus instrument and 10% for confocal microscope....

  20. Wave scattering from statistically rough surfaces

    Bass, F G; ter Haar, D

    2013-01-01

    Wave Scattering from Statistically Rough Surfaces discusses the complications in radio physics and hydro-acoustics in relation to wave transmission under settings seen in nature. Some of the topics that are covered include radar and sonar, the effect of variations in topographic relief or ocean waves on the transmission of radio and sound waves, the reproduction of radio waves from the lower layers of the ionosphere, and the oscillations of signals within the earth-ionosphere waveguide. The book begins with some fundamental idea of wave transmission theory and the theory of random processes a

  1. Surface correlations of hydrodynamic drag for transitionally rough engineering surfaces

    Thakkar, Manan; Busse, Angela; Sandham, Neil

    2017-02-01

    Rough surfaces are usually characterised by a single equivalent sand-grain roughness height scale that typically needs to be determined from laboratory experiments. Recently, this method has been complemented by a direct numerical simulation approach, whereby representative surfaces can be scanned and the roughness effects computed over a range of Reynolds number. This development raises the prospect over the coming years of having enough data for different types of rough surfaces to be able to relate surface characteristics to roughness effects, such as the roughness function that quantifies the downward displacement of the logarithmic law of the wall. In the present contribution, we use simulation data for 17 irregular surfaces at the same friction Reynolds number, for which they are in the transitionally rough regime. All surfaces are scaled to the same physical roughness height. Mean streamwise velocity profiles show a wide range of roughness function values, while the velocity defect profiles show a good collapse. Profile peaks of the turbulent kinetic energy also vary depending on the surface. We then consider which surface properties are important and how new properties can be incorporated into an empirical model, the accuracy of which can then be tested. Optimised models with several roughness parameters are systematically developed for the roughness function and profile peak turbulent kinetic energy. In determining the roughness function, besides the known parameters of solidity (or frontal area ratio) and skewness, it is shown that the streamwise correlation length and the root-mean-square roughness height are also significant. The peak turbulent kinetic energy is determined by the skewness and root-mean-square roughness height, along with the mean forward-facing surface angle and spanwise effective slope. The results suggest feasibility of relating rough-wall flow properties (throughout the range from hydrodynamically smooth to fully rough) to surface

  2. Urban roughness mapping validation techniques and some first results

    Bottema, M; Mestayer, PG

    1998-01-01

    Because of measuring problems related to evaluation of urban roughness parameters, a new approach using a roughness mapping tool has been tested: evaluation of roughness length z(o) and zero displacement z(d) from cadastral databases. Special attention needs to be given to the validation of the

  3. Procedure and applications of combined wheel/rail roughness measurement

    Dittrich, M.G.

    2009-01-01

    Wheel-rail roughness is known to be the main excitation source of railway rolling noise. Besides the already standardised method for direct roughness measurement, it is also possible to measure combined wheel-rail roughness from vertical railhead vibration during a train pass-by. This is a different

  4. Rough sets applied in sublattices and ideals of lattices

    R. Ameri

    2015-12-01

    Full Text Available The purpose of this paper is the study of rough hyperlattice. In this regards we introduce rough sublattice and rough ideals of lattices. We will proceed by obtaining lower and upper approximations in these lattices.

  5. Use of roughness maps in visualisation of surfaces

    Seitavuopio, Paulus; Rantanen, Jukka; Yliruusi, Jouko

    2005-01-01

    monohydrate, theophylline anhydrate, sodium chloride and potassium chloride. The roughness determinations were made by a laser profilometer. The new matrix method gives detailed roughness maps, which are able to show local variations in surface roughness values and provide an illustrative picture...

  6. ROMI 4.0: Updated Rough Mill Simulator

    Timo Grueneberg; R. Edward Thomas; Urs Buehlmann

    2012-01-01

    In the secondary hardwood industry, rough mills convert hardwood lumber into dimension parts for furniture, cabinets, and other wood products. ROMI 4.0, the US Department of Agriculture Forest Service's ROugh-MIll simulator, is a software package designed to simulate the cut-up of hardwood lumber in rough mills in such a way that a maximum possible component yield...

  7. Rough sets selected methods and applications in management and engineering

    Peters, Georg; Ślęzak, Dominik; Yao, Yiyu

    2012-01-01

    Introduced in the early 1980s, Rough Set Theory has become an important part of soft computing in the last 25 years. This book provides a practical, context-based analysis of rough set theory, with each chapter exploring a real-world application of Rough Sets.

  8. 7 CFR 868.201 - Definition of rough rice.

    2010-01-01

    ... 7 Agriculture 7 2010-01-01 2010-01-01 false Definition of rough rice. 868.201 Section 868.201... FOR CERTAIN AGRICULTURAL COMMODITIES United States Standards for Rough Rice Terms Defined § 868.201 Definition of rough rice. Rice (Oryza sativa L.) which consists of 50 percent or more of paddy kernels (see...

  9. Rough viscoelastic sliding contact: Theory and experiments

    Carbone, G.; Putignano, C.

    2014-03-01

    In this paper, we show how the numerical theory introduced by the authors [Carbone and Putignano, J. Mech. Phys. Solids 61, 1822 (2013), 10.1016/j.jmps.2013.03.005] can be effectively employed to study the contact between viscoelastic rough solids. The huge numerical complexity is successfully faced up by employing the adaptive nonuniform mesh developed by the authors in Putignano et al. [J. Mech. Phys. Solids 60, 973 (2012), 10.1016/j.jmps.2012.01.006]. Results mark the importance of accounting for viscoelastic effects to correctly simulate the sliding rough contact. In detail, attention is, first, paid to evaluate the viscoelastic dissipation, i.e., the viscoelastic friction. Fixed the sliding speed and the normal load, friction is completely determined. Furthermore, since the methodology employed in the work allows to study contact between real materials, a comparison between experimental outcomes and numerical prediction in terms of viscoelastic friction is shown. The good agreement seems to validate—at least partially—the presented methodology. Finally, it is shown that viscoelasticity entails not only the dissipative effects previously outlined, but is also strictly related to the anisotropy of the contact solution. Indeed, a marked anisotropy is present in the contact region, which results stretched in the direction perpendicular to the sliding speed. In the paper, the anisotropy of the deformed surface and of the contact area is investigated and quantified.

  10. Roughness as classicality indicator of a quantum state

    Lemos, Humberto C. F.; Almeida, Alexandre C. L.; Amaral, Barbara; Oliveira, Adélcio C.

    2018-03-01

    We define a new quantifier of classicality for a quantum state, the Roughness, which is given by the L2 (R2) distance between Wigner and Husimi functions. We show that the Roughness is bounded and therefore it is a useful tool for comparison between different quantum states for single bosonic systems. The state classification via the Roughness is not binary, but rather it is continuous in the interval [ 0 , 1 ], being the state more classic as the Roughness approaches to zero, and more quantum when it is closer to the unity. The Roughness is maximum for Fock states when its number of photons is arbitrarily large, and also for squeezed states at the maximum compression limit. On the other hand, the Roughness approaches its minimum value for thermal states at infinite temperature and, more generally, for infinite entropy states. The Roughness of a coherent state is slightly below one half, so we may say that it is more a classical state than a quantum one. Another important result is that the Roughness performs well for discriminating both pure and mixed states. Since the Roughness measures the inherent quantumness of a state, we propose another function, the Dynamic Distance Measure (DDM), which is suitable for measure how much quantum is a dynamics. Using DDM, we studied the quartic oscillator, and we observed that there is a certain complementarity between dynamics and state, i.e. when dynamics becomes more quantum, the Roughness of the state decreases, while the Roughness grows as the dynamics becomes less quantum.

  11. Effect of Blade Roughness on Transition and Wind Turbine Performance.

    Ehrmann, Robert S. [Texas A & M Univ., College Station, TX (United States); White, E. B. [Texas A & M Univ., College Station, TX (United States)

    2015-09-01

    The real-world effect of accumulated surface roughness on wind-turbine power production is not well understood. To isolate specific blade roughness features and test their effect, field measurements of turbine-blade roughness were made and simulated on a NACA 633-418 airfoil in a wind tunnel. Insect roughness, paint chips, and erosion were characterized then manufactured. In the tests, these roughness configurations were recreated as distributed roughness, a forward-facing step, and an eroded leading edge. Distributed roughness was tested in three heights and five densities. Chord Reynolds number was varied between 0:8 to 4:8 × 106. Measurements included lift, drag, pitching moment, and boundary-layer transition location. Results indicate minimal effect from paint-chip roughness. As distributed roughness height and density increase, the lift-curve slope, maximum lift, and lift-to-drag ratio decrease. As Reynolds number increases, natural transition is replaced by bypass transition. The critical roughness Reynolds number varies between 178 to 318, within the historical range. At a chord Reynolds number of 3:2 × 106, the maximum lift-to-drag ratio decreases 40% for 140 μm roughness, corresponding to a 2.3% loss in annual energy production. Simulated performance loss compares well to measured performance loss of an in-service wind turbine.

  12. Skin friction measurements of mathematically generated roughness in the transitionally- to fully-rough regimes

    Barros, Julio; Schultz, Michael; Flack, Karen

    2016-11-01

    Engineering systems are affected by surface roughness which cause an increase in drag leading to significant performance penalties. One important question is how to predict frictional drag purely based upon surface topography. Although significant progress has been made in recent years, this has proven to be challenging. The present work takes a systematic approach by generating surface roughness in which surfaces parameters, such as rms , skewness, can be controlled. Surfaces were produced using the random Fourier modes method with enforced power-law spectral slopes. The surfaces were manufactured using high resolution 3D-printing. In this study three surfaces with constant amplitude and varying slope, P, were investigated (P = - 0 . 5 , - 1 . 0 , - 1 . 5). Skin-friction measurements were conducted in a high Reynolds number turbulent channel flow facility, covering a wide range of Reynolds numbers, from hydraulic-smooth to fully-rough regimes. Results show that some long wavelength roughness scales do not contribute significantly to the frictional drag, thus highlighting the need for filtering in the calculation of surface statistics. Upon high-pass filtering, it was found that krms is highly correlated with the measured ks.

  13. Physicochemical characterization of solidification agents used and products formed with radioactive wastes at LWR nuclear power plants

    Kibbey, A.H.; Godbee, H.W.

    1978-01-01

    Solidification of evaporator concentrates, filter sludges, and spent ion exchange resins used in LWR streams is discussed. The introduction of solidification agents to immobilize these sludges and resins can increase the volume of these wastes by a factor of slightly over 1 to greater than 2, depending on the binder chosen. The agents and methods used or proposed for use in solidification of LWR power plant wastes are generally suitable for treating most of the other-than-high-level wastes generated throughout the entire fuel cycle. Among the solidification agents most commonly used or suggested for use are the inorganic cements and organic plastics, which are listed and compared. A summary of considerations important in choosing a solidification agent is presented tabularly

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

    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

  15. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel

  16. Legal, institutional, and political issues in transportation of nuclear materials at the back end of the LWR nuclear fuel cycle

    Lippek, H.E.; Schuller, C.R.

    1979-03-01

    A study was conducted to identify major legal and institutional problems and issues in the transportation of spent fuel and associated processing wastes at the back end of the LWR nuclear fuel cycle. (Most of the discussion centers on the transportation of spent fuel, since this activity will involve virtually all of the legal and institutional problems likely to be encountered in moving waste materials, as well.) Actions or approaches that might be pursued to resolve the problems identified in the analysis are suggested. Two scenarios for the industrial-scale transportation of spent fuel and radioactive wastes, taken together, high-light most of the major problems and issues of a legal and institutional nature that are likely to arise: (1) utilizing the Allied General Nuclear Services (AGNS) facility at Barnwell, SC, as a temporary storage facility for spent fuel; and (2) utilizing AGNS for full-scale commercial reprocessing of spent LWR fuel.

  17. Parameters' influence estimation on Puf supply and demand in transitional period from LWR to FBR in Japan

    Kobayashi, Hiroaki; Ohta, Hirokazu; Inoue, Tadashi

    2009-01-01

    Plutonium fissile (Puf) amounts to balance supply and demand during transition period were evaluated with different parameters. Estimated total Puf demand in transitional period was sensitive to deployment speed of FBR. Because FBRs will be deployed as replacements of old LWRs for keeping total capacity, deployment history of existing LWRs should be taken into consideration. According to the estimation, LWR fuel burnup and utilized capacity are not big issue. Because certain amount of LWR spent fuel will remain in early phase of transitional period, there is enough time for preparing Puf supply. On the other hand, FBR fuel cycle time (SF cooling time + fuel fabrication time) have large impact on Puf supply. Fuel cycle technologies including transportation for applying to short cooling spent fuels should be developed. (author)

  18. Influence of surface roughness of a desert

    Sud, Y. C.; Smith, W. E.

    1984-01-01

    A numerical simulation study, using the current GLAS climate GCM, was carried out to examine the influence of low bulk aerodynamic drag parameter in the deserts. The results illustrate the importance of yet another feedback effect of a desert on itself, that is produced by the reduction in surface roughness height of land once the vegetation dies and desert forms. Apart from affecting the moisture convergence, low bulk transport coefficients of a desert lead to enhanced longwave cooling and sinking which together reduce precipitation by Charney's (1975) mechanism. Thus, this effect, together with albedo and soil moisture influence, perpetuate a desert condition through its geophysical feedback effect. The study further suggests that man made deserts is a viable hypothesis.

  19. Identification of the impacts of maintenance and testing upon the safety of LWR power plants. Part II. Final report

    Husseiny, A.A.; Sabri, Z.A.; Turnage, J.J.

    1980-04-01

    Information is presented concerning overview of literature relating to radiation exposure and operating experience; details of LWR-MTC3 classification system; histograms for individual BWR facilities depicting frequency of M and T mode and frequency of systems and components involved with M and T problems; histograms for individual PWR facilities depicting frequency of M and T mode and frequency of systems and components involved with M and T problems; and Fortran program for M and T data clustering

  20. Evaluation of conceptual flowsheets for incorporating Light Water Reactor (LWR) fuel materials in an advanced nuclear fuel cycle

    Bell, J.T.; Burch, W.D.; Collins, E.D.; Forsberg, C.W.; Prince, B.E.; Bond, W.D.; Campbell, D.O.; Delene, J.G.; Mailen, J.C.

    1990-08-01

    A preliminary study by a group of experts at ORNL has generated and evaluated a number of aqueous and non-aqueous flowsheets for recovering transuranium actinides from LWR fuel for use as fuel in an LMR and, at the same time, for transmutation of the wastes to less hazardous materials. The need for proliferation resistance was a consideration in the flowsheets. The current state of development of the flowsheets was evaluated and recommendations for additional study were made. 3 refs., 6 figs

  1. Minutes of the 13th light water reactor pressure vessel surveillance dosimetry improvement program (LWR-PV-SDIP) meeting

    1984-04-01

    Information is presented concerning ASTM LWR standards and program documentation; trend curves, PSF, and other test reactor metallurgical programs; PSF dosimetry and metallurgical capsule neutron and gamma environment characterization and metallurgical studies; PVS characterization program; other neutron fields; surveillance dosimetry measurement facility (SDMF) and perturbation studies; transport theory calculations; gamma field benchmarks and photo-reaction studies; and fission and non-fission sensor inventories and quality assurance

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

    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)

  3. Fast reactor core design studies to cope with TRU fuel composition changes in the LWR-to-FBR transition period

    Kawashima, Katsuyuki; Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

    2009-01-01

    As part of the Fast Reactor Cycle Technology Development Project (FaCT Project), sodium-cooled fast reactor core design efforts have been made to cope with the TRU fuel composition changes expected during LWR-to-FBR transition period, in which a various kind of TRU fuel compositions are available depending on the characteristics of the LWR spent fuels and a way of recycling them. A 750 MWe mixed-oxide fuel core is firstly defined as a FaCT medium-size reference core and its neutronics characteristics are determined. The core is a high internal conversion type and has an average burnup of 150 GWD/T. The reference TRU fuel composition is assumed to come from the FBR equilibrium state. Compared to the LWR-to-FBR transition period, the TRU fuels in the FBR equilibrium period are multi-recycled through fast reactors and have a different composition. An available TRU fuel composition is determined by fast reactor spent fuel multi-recycling scenarios. Then the FaCT core corresponding to the TRU fuel with different compositions is set according to the TRU fuel composition changes in LWR-to-FBR transition period, and the key core neutronics characteristics are assessed. It is shown that among the core neutronics characteristics, the burnup reactivity and the safety parameters such as sodium void reactivity and Doppler coefficient are significantly influenced by the TRU fuel composition changes. As a result, a general characteristic in the FaCT core design to cope with TRU fuel composition changes is grasped and the design envelopes are identified in terms of the burnup reactivity and the safety parameters. (author)

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

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

    2010-06-01

    Availability of enough cooling water has been one of the major issues for the nuclear power plant site selection. Cooling water issues have frequently disrupted the normal operation at some nuclear power plants during heat waves and long draught. The issues become more severe due to the new round of nuclear power expansion and global warming. During hot summer days, cooling water leaving a power plant may become too hot to threaten aquatic life so that environmental regulations may force the plant to reduce power output or even temporarily to be shutdown. For new nuclear power plants to be built at areas without enough cooling water, dry cooling can be used to remove waste heat directly into the atmosphere. However, dry cooling will result in much lower thermal efficiency when the weather is hot. One potential solution for the above mentioned issues is to use ice thermal storage systems (ITS) that reduce cooling water requirements and boost the plant’s thermal efficiency in hot hours. ITS uses cheap off-peak electricity to make ice and uses those ice for supplemental cooling during peak demand time. ITS is suitable for supplemental cooling storage due to its very high energy storage density. ITS also provides a way to shift large amount of electricity from off peak time to peak time. Some gas turbine plants already use ITS to increase thermal efficiency during peak hours in summer. ITSs have also been widely used for building cooling to save energy cost. Among three cooling methods for LWR applications: once-through, wet cooling tower, and dry cooling tower, once-through cooling plants near a large water body like an ocean or a large lake and wet cooling plants can maintain the designed turbine backpressure (or condensation temperature) during 99% of the time; therefore, adding ITS to those plants will not generate large benefits. For once-through cooling plants near a limited water body like a river or a small lake, adding ITS can bring significant economic

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

    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. Stylized whole-core benchmark of the Integral Inherently Safe Light Water Reactor (I2S-LWR) concept

    Hon, Ryan; Kooreman, Gabriel; Rahnema, Farzad; Petrovic, Bojan

    2017-01-01

    Highlights: • A stylized benchmark specification of the I2S-LWR core. • A library of cross sections were generated in both 8 and 47 groups. • Monte Carlo solutions generated for the 8 group library using MCNP5. • Cross sections and pin fission densities provided in journal’s repository. - Abstract: The Integral, Inherently Safe Light Water Reactor (I 2 S-LWR) is a pressurized water reactor (PWR) concept under development by a multi-institutional team led by Georgia Tech. The core is similar in size to small 2-loop PWRs while having the power level of current large reactors (∼1000 MWe) but using uranium silicide fuel and advanced stainless steel cladding. A stylized benchmark specification of the I 2 S-LWR core has been developed in order to test whole-core neutronics codes and methods. For simplification the core was split into 57 distinct material regions for cross section generation. Cross sections were generated using the lattice physics code HELIOS version 1.10 in both 8 and 47 groups. Monte Carlo solutions, including eigenvalue and pin fission densities, were generated for the 8 group library using MCNP5. Due to space limitations in this paper, the full cross section library and normalized pin fission density results are provided in the journal’s electronic repository.

  7. Testing round robin on cyclic crack growth of low and medium sulfur A533-B steels in LWR environments

    Kitagawa, H.; Komai, K.; Nakajima, H.; Higuchi, M.

    1987-01-01

    After the facts of environmentally assisted crack growth of low alloy steel was first observed when cyclically loaded in high temperature water. The subject has been extensively studied in connection with the evaluation of the integrity of LWR pressure boundary materials. In 1977, International Cooperative Group on Cyclic Crack Growth Rate Testing Evaluation (the ICCGR group) was organized for more systematic and effective solution of the problem. Successful results have been reported on the programs of the ICCGR activity, particularly in the promotion of a couple of programs of testing round robin and the associated research. JAERI also organized a domestic group of 15 organizations as the Corrosion Fatigue Subcommittee(JCF) of the LWR Safety Research Committee to carry out the similar test program. The group has been evaluating the behavior of steels representing the range of quality for the existing Japanese LWR plants. This paper describes the present status of the Japanese domestic testing round robin and related research especially focused on the test methodology

  8. Safety aspects of LWR fuel reprocessing and mixed oxide fuel fabrication plants

    Fischer, M.; Leichsenring, C.H.; Herrmann, G.W.; Schueller, W.; Hagenberg, W.; Stoll, W.

    1977-01-01

    The paper is focused on the safety and the control of the consequences of credible accidents in LWR fuel reprocessing plants and in mixed oxide fuel fabrication plants. Each of these plants serve for many power reactor (about 50.000 Mwel) thus the contribution to the overall risk of nuclear energy is correspondingly low. Because of basic functional differences between reprocessing plants, fuel fabrication plants and nuclear power reactors, the structure and safety systems of these plants are different in many respects. The most important differences that influence safety systems are: (1) Both fuel reprocessing and fabrication plants do not have the high system pressure that is associated with power reactors. (2) A considerable amount of the radioactivity of the fuel, which is in the form of short-lived radionuclides has decayed. Therefore, fuel reprocessing plants and mixed oxide fuel fabrication plants are designed with multiple confinement barriers for control of radioactive materials, but do not require the high-pressure containment systems that are used in LWR plants. The consequences of accidents which may lead to the dispersion of radioactive materials such as chemical explosions, nuclear excursions, fires and failure of cooling systems are considered. A reasonable high reliability of the multiple confinement approach can be assured by design. In fuel reprocessing plants, forced cooling is necessary only in systems where fission products are accumulated. However, the control of radioactive materials can be maintained during normal operation and during the above mentioned accidents, if the dissolver off-gas and vessel off-gas treatment systems provide for effective removal of radioactive iodine, radioactive particulates, nitrogen oxides, tritium and krypton 85. In addition, the following incidents in the dissolver off-gas system itself must be controlled: failures of iodine filters, hydrogen explosion in O 2 - and NOsub(x)-reduction component, decomposition of

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

    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

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

    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

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

    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

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

    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

  13. Urban Aerodynamic Roughness Length Mapping Using Multitemporal SAR Data

    Fengli Zhang

    2017-01-01

    Full Text Available Aerodynamic roughness is very important to urban meteorological and climate studies. Radar remote sensing is considered to be an effective means for aerodynamic roughness retrieval because radar backscattering is sensitive to the surface roughness and geometric structure of a given target. In this paper, a methodology for aerodynamic roughness length estimation using SAR data in urban areas is introduced. The scale and orientation characteristics of backscattering of various targets in urban areas were firstly extracted and analyzed, which showed great potential of SAR data for urban roughness elements characterization. Then the ground truth aerodynamic roughness was calculated from wind gradient data acquired by the meteorological tower using fitting and iterative method. And then the optimal dimension of the upwind sector for the aerodynamic roughness calculation was determined through a correlation analysis between backscattering extracted from SAR data at various upwind sector areas and the aerodynamic roughness calculated from the meteorological tower data. Finally a quantitative relationship was set up to retrieve the aerodynamic roughness length from SAR data. Experiments based on ALOS PALSAR and COSMO-SkyMed data from 2006 to 2011 prove that the proposed methodology can provide accurate roughness length estimations for the spatial and temporal analysis of urban surface.

  14. Roughness coefficient and its uncertainty in gravel-bed river

    Ji-Sung Kim

    2010-06-01

    Full Text Available Manning's roughness coefficient was estimated for a gravel-bed river reach using field measurements of water level and discharge, and the applicability of various methods used for estimation of the roughness coefficient was evaluated. Results show that the roughness coefficient tends to decrease with increasing discharge and water depth, and over a certain range it appears to remain constant. Comparison of roughness coefficients calculated by field measurement data with those estimated by other methods shows that, although the field-measured values provide approximate roughness coefficients for relatively large discharge, there seems to be rather high uncertainty due to the difference in resultant values. For this reason, uncertainty related to the roughness coefficient was analyzed in terms of change in computed variables. On average, a 20% increase of the roughness coefficient causes a 7% increase in the water depth and an 8% decrease in velocity, but there may be about a 15% increase in the water depth and an equivalent decrease in velocity for certain cross-sections in the study reach. Finally, the validity of estimated roughness coefficient based on field measurements was examined. A 10% error in discharge measurement may lead to more than 10% uncertainty in roughness coefficient estimation, but corresponding uncertainty in computed water depth and velocity is reduced to approximately 5%. Conversely, the necessity for roughness coefficient estimation by field measurement is confirmed.

  15. Rough Electrode Creates Excess Capacitance in Thin-Film Capacitors.

    Torabi, Solmaz; Cherry, Megan; Duijnstee, Elisabeth A; Le Corre, Vincent M; Qiu, Li; Hummelen, Jan C; Palasantzas, George; Koster, L Jan Anton

    2017-08-16

    The parallel-plate capacitor equation is widely used in contemporary material research for nanoscale applications and nanoelectronics. To apply this equation, flat and smooth electrodes are assumed for a capacitor. This essential assumption is often violated for thin-film capacitors because the formation of nanoscale roughness at the electrode interface is very probable for thin films grown via common deposition methods. In this work, we experimentally and theoretically show that the electrical capacitance of thin-film capacitors with realistic interface roughness is significantly larger than the value predicted by the parallel-plate capacitor equation. The degree of the deviation depends on the strength of the roughness, which is described by three roughness parameters for a self-affine fractal surface. By applying an extended parallel-plate capacitor equation that includes the roughness parameters of the electrode, we are able to calculate the excess capacitance of the electrode with weak roughness. Moreover, we introduce the roughness parameter limits for which the simple parallel-plate capacitor equation is sufficiently accurate for capacitors with one rough electrode. Our results imply that the interface roughness beyond the proposed limits cannot be dismissed unless the independence of the capacitance from the interface roughness is experimentally demonstrated. The practical protocols suggested in our work for the reliable use of the parallel-plate capacitor equation can be applied as general guidelines in various fields of interest.

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

    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

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

    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

  18. Narrow line width operation of a 980 nm gain guided tapered diode laser bar

    Vijayakumar, Deepak; Jensen, Ole Bjarlin; Barrientos-Barria, Jessica

    2011-01-01

    We demonstrate two different schemes for the spectral narrowing of a 12 emitter 980 nm gain guided tapered diode laser bar. In the first scheme, a reflective grating has been used in a Littman Metcalf configuration and the wavelength of the laser emission could be narrowed down from more than 5.......5 nm in the free running mode to 0.04 nm (FWHM) at an operating current of 30 A with an output power of 8 W. The spectrum was found to be tunable within a range of 16 nm. In the second scheme, a volume Bragg grating has been used to narrow the wavelength of the laser bar from over 5 nm to less than 0.......2 nm with an output of 5 W at 20 A. To our knowledge, this is the first time spectral narrowing has been performed on a gain guided tapered diode laser bar. In the Littman Metcalf configuration, the spectral brightness has been increased by 86 times and in the volume Bragg grating cavity the spectral...

  19. Recrystallization curve study of zircaloy-4 with DRX line width method

    Juarez, G; Buioli, C; Samper, R; Vizcaino, P

    2012-01-01

    X-ray diffraction peak broadening analysis is a method that allows to characterize the plastic deformation in metals. This technique is a complement of transmission electron microscopy (TEM) to determine dislocation densities. So that, both techniques may cover a wide range in the analysis of metals deformation. The study of zirconium alloys is an issue of usual interest in the nuclear industry since such materials present the best combination of good mechanical properties, corrosion behavior and low neutron cross section. It is worth noting there are two factors to be taken into account in the application of the method developed for this purpose: the characteristic anisotropy of the hexagonals and the strong texture that these alloys acquire during the manufacturing process. In order to assess the recrystallization curve of Zircaloy-4, a powder of this alloy was produced through filing. Then, fractions of the powder were subjected to thermal treatments at different temperatures for the same time. Since the powder has a random crystallographic orientation, the texture effect practically disappears; this is the reason why the Williamson and Hall method may be easily used, producing good fittings and predicting confidence values of diffraction domain size and the accumulated deformation. The temperatures selected for the thermal treatments were 1000, 700, 600, 500, 420, 300 and 200 o C during 2 hs. As a result of these annealings, powders in different recovery stages were obtained (completely recrystallized, partially recrystallized and non-recrystallized structures with different levels of stress relieve). The obtained values were also compared with the non annealed powder ones. All the microstructural evolution through the annealings was followed by optical microscopy (author)

  20. Distributed seeding for narrow-line width hard x-ray free-electron lasers

    Nguyen, Dinh Cong [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anisimov, Petr Mikhaylovich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lewellen, IV, John W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Marksteiner, Quinn R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-09

    We describe a new FEL line-narrowing technique called distributed seeding (DS), using Si(111) Bragg crystal monochromators to enhance the spectral brightness of the MaRIE hard X-ray freeelectron laser. DS differs from self-seeding in three important aspects. First, DS relies on spectral filtering of the radiation at multiple locations along the undulator, with a monochromator located every few power gain lengths. Second, DS performs filtering early in the exponential gain region before SASE spikes start to appear in the radiation longitudinal profile. Third, DS provides the option to select a wavelength longer than the peak of the SASE gain curve, which leads to improved spectral contrast of the seeded FEL over the SASE background. Timedependent Genesis simulations show the power-vs-z growth curves for DS exhibit behaviors of a seeded FEL amplifier, such as exponential growth region immediately after the filters. Of the seeding approaches considered, the two-stage DS spectra produce the highest contrast of seeded FEL over the SASE background and that the three-stage DS provides the narrowest linewidth with a relative spectral FWHM of 8 X 10-5 .

  1. Spectroscopic Studies of Solar Corona VI: Trend in Line-width ...

    cm coronagraph at the Norikura Solar Observatory on several days during the years 1997–2004. The Coude type .... Top-most panel shows the variation of FWHM of the 6374 Е and 5303 Е emission lines with height above the limb when all ...

  2. A simple method for controlling the line width of SASE X-ray FELs

    Geloni, Gianluca; Kocharyan, Vitali; Saldin, Evgeni

    2010-04-01

    This paper describes a novel single-bunch self-seeding scheme for generating highly monochromatic X-rays from a baseline XFEL undulator. A self-seeded XFEL consists of two undulators with an X-ray monochromator located between them. Previous self-seeding schemes made use of a four-crystal fixed-exit monochromator in Bragg geometry. In such monochromator the X-ray pulse acquires a cm-long path delay, which must be compensated. For a single-bunch self-seeding scheme this requires a long electron beam bypass, implying modifications of the baseline undulator configuration. To avoid this problem, a double bunch self-seeding scheme based on a special photoinjector setup was recently proposed. At variance, here we propose a new time-domain method of monochromatization exploiting a single crystal in the transmission direction, thus avoiding the problem of extra-path delay for the X-ray pulse. The method can be realized using a temporal windowing technique, requiring a magnetic delay for the electron bunch only. When the incident X-ray beam satisfies the Bragg diffraction condition, multiple scattering takes place and the transmittance spectrum in the crystal exhibits an absorption resonance with a narrow linewidth. Then, the temporal waveform of the transmitted radiation pulse is characterized by a long monochromatic wake. The radiation power within this wake is much larger than the shot noise power. At the entrance of the second undulator, the monochromatic wake of the radiation pulse is combined with the delayed electron bunch, and amplified up to saturation level. The proposed setup is extremely simple and composed of as few as two simple elements. These are the crystal and the short magnetic chicane, which accomplishes three tasks by itself. It creates an offset for crystal installation, it removes the electron micro-bunching produced in the first undulator, and it acts as a delay line for temporal windowing. Using a single crystal installed within a short magnetic chicane in the baseline undulator, it is possible to decrease the bandwidth of the radiation well beyond the XFEL design down to 10 -5 . The installation of the magnetic chicane does not perturb the undulator focusing system and does not interfere with the baseline mode of operation. We present feasibility study and exemplifications for the SASE2 line of the European XFEL. (orig.)

  3. InAlGaAs/AlGaAs quantum wells: line widths, transition energies and segregation

    Jensen, Jacob Riis; Hvam, Jørn Märcher; Langbein, Wolfgang

    2000-01-01

    We investigate the optical properties of InAlCaAs/AlGaAs quantum wells pseudomorphically grown on GaAs using molecular beam epitaxy (MBE). The transition energies, measured with photoluminescence (PL), are modelled solving the Schrodinger equation, and taking into account segregation in the group...

  4. Methane spectral line widths and shifts, and dependences on physical parameters

    Fox, K.; Quillen, D. T.; Jennings, D. E.; Wagner, J.; Plymate, C.

    1991-01-01

    A detailed report of the recent high-resolution spectroscopic research on widths and shifts measured for a strong infrared-active fundamental of methane is presented. They were measured in collision with several rare gases and diatomic molecules, in the vibrational-rotational fundamental near 3000/cm. These measurements were made at an ambient temperature of 294 K over a range of pressures from 100 to 700 torr. The measurements are discussed in a preliminary but detailed and quantitative manner with reference to masses, polarizabilities, and quadrupole moments. Some functional dependences on these physical parameters are considered. The present data are useful for studies of corresponding planetary spectra.

  5. Tunable, Narrow Line Width Mid-Infrared Laser Source, Phase II

    National Aeronautics and Space Administration — The purpose of this project is to advance the technology of interband cascade (IC) lasers and their facet coatings and to design, build, and deliver to NASA a...

  6. Tunable, Narrow Line Width Mid-Infrared Laser Source, Phase I

    National Aeronautics and Space Administration — Maxion Technologies, Inc. (Maxion) and Professor Mario Dagenais and his group at the University of Maryland (UMD) jointly propose to develop a compact, efficient,...

  7. Multi-scale analysis of the roughness effect on lubricated rough contact

    DEMIRCI, Ibrahim

    2014-01-01

    Determining friction is as equally essential as determining the film thickness in the lubricated contact, and is an important research subject. Indeed, reduction of friction in the automotive industry is important for both the minimization of fuel consumption as well as the decrease in the emissions of greenhouse gases. However, the progress in friction reduction has been limited by the difficulty in understanding the mechanism of roughness effects on friction. It was observed that micro-surf...

  8. Rough surface mitigates electron and gas emission

    Molvik, A.

    2004-01-01

    Heavy-ion beams impinging on surfaces near grazing incidence (to simulate the loss of halo ions) generate copious amounts of electrons and gas that can degrade the beam. We measured emission coefficients of η e (le) 130 and η 0 ∼ 10 4 respectively, with 1 MeV K + incident on stainless steel. Electron emission scales as η e ∝ 1/cos(θ), where θ is the ion angle of incidence relative to normal. If we were to roughen a surface by blasting it with glass beads, then ions that were near grazing incidence (90 o ) on smooth surface would strike the rims of the micro-craters at angles closer to normal incidence. This should reduce the electron emission: the factor of 10 reduction, Fig. 1(a), implies an average angle of incidence of 62 o . Gas desorption varies more slowly with θ (Fig. 1(b)) decreasing a factor of ∼2, and along with the electron emission is independent of the angle of incidence on a rough surface. In a quadrupole magnet, electrons emitted by lost primary ions are trapped near the wall by the magnetic field, but grazing incidence ions can backscatter and strike the wall a second time at an azimuth where magnetic field lines intercept the beam. Then, electrons can exist throughout the beam (see the simulations of Cohen, HIF News 1-2/04). The SRIM (TRIM) Monte Carlo code predicts that 60-70% of 1 MeV K + ions backscatter when incident at 88-89 o from normal on a smooth surface. The scattered ions are mostly within ∼10 o of the initial direction but a few scatter by up to 90 o . Ion scattering decreases rapidly away from grazing incidence, Fig. 1(c ). At 62 deg. the predicted ion backscattering (from a rough surface) is 3%, down a factor of 20 from the peak, which should significantly reduce electrons in the beam from lost halo ions. These results are published in Phys. Rev. ST - Accelerators and Beams

  9. Influence of edge roughness on graphene nanoribbon resonant tunnelling diodes

    Liang Gengchiau; Khalid, Sharjeel Bin; Lam, Kai-Tak

    2010-01-01

    The edge roughness effects of graphene nanoribbons on their application in resonant tunnelling diodes with different geometrical shapes (S, H and W) were investigated. Sixty samples for each 5%, 10% and 15% edge roughness conditions of these differently shaped graphene nanoribbon resonant tunnelling diodes were randomly generated and studied. Firstly, it was observed that edge roughness in the barrier regions decreases the effective barrier height and thickness, which increases the broadening of the quantized states in the quantum well due to the enhanced penetration of the wave-function tail from the electrodes. Secondly, edge roughness increases the effective width of the quantum well and causes the lowering of the quantized states. Furthermore, the shape effects on carrier transport are modified by edge roughness due to different interfacial scattering. Finally, with the effects mentioned above, edge roughness has a considerable impact on the device performance in terms of varying the peak-current positions and degrading the peak-to-valley current ratio.

  10. A new fiber optic sensor for inner surface roughness measurement

    Xu, Xiaomei; Liu, Shoubin; Hu, Hong

    2009-11-01

    In order to measure inner surface roughness of small holes nondestructively, a new fiber optic sensor is researched and developed. Firstly, a new model for surface roughness measurement is proposed, which is based on intensity-modulated fiber optic sensors and scattering modeling of rough surfaces. Secondly, a fiber optical measurement system is designed and set up. Under the help of new techniques, the fiber optic sensor can be miniaturized. Furthermore, the use of micro prism makes the light turn 90 degree, so the inner side surface roughness of small holes can be measured. Thirdly, the fiber optic sensor is gauged by standard surface roughness specimens, and a series of measurement experiments have been done. The measurement results are compared with those obtained by TR220 Surface Roughness Instrument and Form Talysurf Laser 635, and validity of the developed fiber optic sensor is verified. Finally, precision and influence factors of the fiber optic sensor are analyzed.

  11. Turbulent flow with suction in smooth and rough pipes

    Verdier, Andre.

    1977-11-01

    It concerns an experimental study of turbulent flow inside a pipe with rough and porous wall and suction applied through it. The first part recall the basic knowledge concerning the turbulent flow with roughness. In second part statistical equations of fluid wall stress are written in the case of a permeable rough wall, in order to underline the respective role played by viscosity and pressure terms. In the third part the dynamic equilibrium of the flow is experimentally undertaken in the smooth and rough range with and without wall suction. Some empirical formulae are proposed for the mean velocity profiles in the inertial range and for friction velocity with suction. In the case of the sand roughness used, it does not seem that critical Reynolds number of transition from smooth to rough range is varied [fr

  12. Numerical simulations of seepage flow in rough single rock fractures

    Qingang Zhang

    2015-09-01

    Full Text Available To investigate the relationship between the structural characteristics and seepage flow behavior of rough single rock fractures, a set of single fracture physical models were produced using the Weierstrass–Mandelbrot functions to test the seepage flow performance. Six single fractures, with various surface roughnesses characterized by fractal dimensions, were built using COMSOL multiphysics software. The fluid flow behavior through the rough fractures and the influences of the rough surfaces on the fluid flow behavior was then monitored. The numerical simulation indicates that there is a linear relationship between the average flow velocity over the entire flow path and the fractal dimension of the rough surface. It is shown that there is good a agreement between the numerical results and the experimental data in terms of the properties of the fluid flowing through the rough single rock fractures.

  13. Effect of truncated cone roughness element density on hydrodynamic drag

    Womack, Kristofer; Schultz, Michael; Meneveau, Charles

    2017-11-01

    An experimental study was conducted on rough-wall, turbulent boundary layer flow with roughness elements whose idealized shape model barnacles that cause hydrodynamic drag in many applications. Varying planform densities of truncated cone roughness elements were investigated. Element densities studied ranged from 10% to 79%. Detailed turbulent boundary layer velocity statistics were recorded with a two-component LDV system on a three-axis traverse. Hydrodynamic roughness length (z0) and skin-friction coefficient (Cf) were determined and compared with the estimates from existing roughness element drag prediction models including Macdonald et al. (1998) and other recent models. The roughness elements used in this work model idealized barnacles, so implications of this data set for ship powering are considered. This research was supported by the Office of Naval Research and by the Department of Defense (DoD) through the National Defense Science & Engineering Graduate Fellowship (NDSEG) Program.

  14. Irregular wall roughness in turbulent Taylor-Couette flow

    Berghout, Pieter; Zhu, Xiaojue; Verzicco, Roberto; Lohse, Detlef; Stevens, Richard

    2017-11-01

    Many wall bounded flows in nature, engineering and transport are affected by surface roughness. Often, this has adverse effects, e.g. drag increase leading to higher energy costs. A major difficulty is the infinite number of roughness geometries, which makes it impossible to systematically investigate all possibilities. Here we present Direct Numerical Simulations (DNS) of turbulent Taylor-Couette flow. We focus on the transitionally rough regime, in which both viscous and pressure forces contribute to the total wall stress. We investigate the effect of the mean roughness height and the effective slope on the roughness function, ΔU+ . Also, we present simulations of varying Ta (Re) numbers for a constant mean roughness height (kmean+). Alongside, we show the behavior of the large scale structures (e.g. plume ejection, Taylor rolls) and flow structures in the vicinity of the wall.

  15. Turbulence modifications in a turbulent boundary layer over a rough wall with spanwise-alternating roughness strips

    Bai, H. L.; Kevin, Hutchins, N.; Monty, J. P.

    2018-05-01

    Turbulence modifications over a rough wall with spanwise-varying roughness are investigated at a moderate Reynolds number Reτ ≈ 2000 (or Reθ ≈ 6400), using particle image velocimetry (PIV) and hotwire anemometry. The rough wall is comprised of spanwise-alternating longitudinal sandpaper strips of two different roughness heights. The ratio of high- and low-roughness heights is 8, and the ratio of high- and low-roughness strip width is 0.5. PIV measurements are conducted in a wall-parallel plane located in the logarithmic region, while hotwire measurements are made throughout the entire boundary layer in a cross-stream plane. In a time-average sense, large-scale counter-rotating roll-modes are observed in the cross-stream plane over the rough wall, with downwash and upwash common-flows displayed over the high- and low-roughness strips, respectively. Meanwhile, elevated and reduced streamwise velocities occur over the high- and low-roughness strips, respectively. Significant modifications in the distributions of mean vorticities and Reynolds stresses are observed, exhibiting features of spatial preference. Furthermore, spatial correlations and conditional average analyses are performed to examine the alterations of turbulence structures over the rough wall, revealing that the time-invariant structures observed are resultant from the time-average process of instantaneous turbulent events that occur mostly and preferentially in space.

  16. Radiation-induced grain subdivision and bubble formation in U3Si2 at LWR temperature

    Yao, Tiankai; Gong, Bowen; He, Lingfeng; Harp, Jason; Tonks, Michael; Lian, Jie

    2018-01-01

    U3Si2, an advanced fuel form proposed for light water reactors (LWRs), has excellent thermal conductivity and a high fissile element density. However, limited understanding of the radiation performance and fission gas behavior of U3Si2 is available at LWR conditions. This study explores the irradiation behavior of U3Si2 by 300 keV Xe+ ion beam bombardment combining with in-situ transmission electron microscopy (TEM) observation. The crystal structure of U3Si2 is stable against radiation-induced amorphization at 350 °C even up to a very high dose of 64 displacements per atom (dpa). Grain subdivision of U3Si2 occurs at a relatively low dose of 0.8 dpa and continues to above 48 dpa, leading to the formation of high-density nanoparticles. Nano-sized Xe gas bubbles prevail at a dose of 24 dpa, and Xe bubble coalescence was identified with the increase of irradiation dose. The volumetric swelling resulting from Xe gas bubble formation and coalescence was estimated with respect to radiation dose, and a 2.2% volumetric swelling was observed for U3Si2 irradiated at 64 dpa. Due to extremely high susceptibility to oxidation, the nano-sized U3Si2 grains upon radiation-induced grain subdivision were oxidized to nanocrystalline UO2 in a high vacuum chamber for TEM observation, eventually leading to the formation of UO2 nanocrystallites stable up to 80 dpa.

  17. Characterization of LWR fuel rod irradiations with power transients in the BR2 reflector

    Ponsard, B.; Bodart, S.; Meer, K. van der; Raedt, C. de

    1996-01-01

    Fuel rod irradiations in reflector positions of the materials testing reactor BR2 are becoming increasingly important. A typical example is that of irradiation devices containing single LWR fuel rods, to be tested in the framework of a new international fuel investigation and development programme. Some of the irradiations will comprise power transients with central fuel melting (at 2800 deg. C), the power increase being obtained by decreasing the pressure in a He-3 neutron absorbing screen and/or by varying the BR2 reactor operating power. A total power variation by a factor of at least 2.5 in the fuel rod irradiated could thus be achieved. In some of the rods, central temperature measurements (up to 2000 deg. C) will be carried out. Both fresh and pre-irradiated fuel rods are concerned in the programme. For these irradiations, the accurate knowledge of the neutron-induced fission heating and of the gamma heating is required, as one of the purposes of the programme consists in establishing the correlation among the thermal conductivity, the burn-up and the irradiation temperature. Calibration work among various measuring methods and between measurements and one- and two-dimensional calculations is being pursued. (author). 10 refs, 15 figs, 3 tabs

  18. The Impact of Fukushima Accidents on LWR Safety and the Nuclear Power Risks

    Sehgal, B. R.

    2014-01-01

    The history of the consideration of severe accidents (SA) safety begins really with WASH-1400 [1] initiated by USNRC in early 1970s. The WASH-1400 considered accidents of decreasing probability and increasing consequence.The accidents considered, occurred due to successive faults which lead to at least the melting of the core and a possible radioactivity release to the environment. The increasing consequence accidents would entail additional failures e.g., vessel failure, late containment failure, containment bypass, early containment failure etc. These additional failures would lead to larger releases of radioactivity and thus larger consequences for the public in the vicinity of the plant. WASH -1400 did not provide estimates of the costs for cleanup of the contaminated land area. Also there were no estimates of the economic costs involved in removal of the molten fuel and the decommissioning of the stricken plant. The emphasis in WASH-1400 was primarily with physical damage to the population in the vicinity of the plant and peripherally with the societal, social and economic costs of a severe accident in a large LWR plant

  19. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L.; Xolocostli M, J. V.; Gomez T, A. M.

    2016-09-01

    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_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_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)

  20. MODRIB - a zero dimensional code for criticality and burn-up of LWR's

    Gaafar, M.A.; El-Cherif, A.I.

    1980-01-01

    The computer program MODRIB is a zero-dimensional code for calculating criticality and burn-up of light water reactors (LWR's). It is a version of an Italian code RIBOT-2 with an updated cross-section data library. The nuclear constants of MODRIB-code are calculated with a two group scheme (fast and thermal), where the fast group is an average of three fast groups. The code requires as input data essential extensive reactor parameters such as fuel rod radius, clad thickness, fuel enrichment, lattice pitch, water density and temperature etc. A summary of the physical model description and the input-output procedures are given in this report. Selected results of two sample problems are also given for the purpose of checking the validity and reliability of the code. The first is BWR and the second is PWR. The calculation time for a criticality problem with burn-up is about 8 seconds for the first time step and about 3 seconds for each subsequent time step on the ICL-1906 computer facility. The requirements on the memory size is less than 32 K-word. (author)

  1. Long-term aging embrittlement of cast duplex stainless steels in LWR systems

    Chopra, O.K.; Chung, H.M.

    1991-01-01

    The primary objectives of this program are to investigate the significance of in-service embrittlement of cast duplex stainless steels in light water reactor (LWR) systems and to evaluate possible remedies for the embrittlement problem in existing and future plants. The scope of the investigation includes three goals: (1) develop a methodology and correlations for predicting the toughness loss suffered by cast stainless steel components during normal and extended life of LWRs, (2) validate the simulation of in-reactor degradation by accelerated aging, and (3) establish the effects of key compositional and metallurgical variables on the kinetics and extent of embrittlement. The emphasis during the current year was on developing a procedure and correlations for predicting fracture toughness J-R curves of aged cast stainless steels from known material information. The present analysis has focused on developing correlations for the fracture properties in terms of material information that can be determined from the certified material test record (CMTR) and on ensuring that the correlations are adequately conservative for structurally weak materials

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

    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)

  3. Analysis of LWR Full MOX Core Physics Experiments with Major Nuclear Data Libraries

    Yamamoto, Toru [Japan Nuclear Energy Safety Organization, Tokyo (Japan)

    2007-07-01

    Nuclear Power Engineering Corporation (NUPEC) studied high moderation full MOX cores as a part of advanced LWR core concept studies from 1994 to 2003 supported by the Ministry of Economy, Trade and Industry. In order to obtain the major physics characteristics of such advanced MOX cores, NUPEC carried out core physics experimental programs called MISTRAL and BASALA from 1996 to 2002 in the EOLE critical facility of the Cadarache Center in collaboration with CEA. NUPEC also obtained a part of experimental data of the EPICURE program that CEA had conducted for 30 % Pu recycling in French PWRs. Japan Nuclear Energy Safety Organization(JNES) established in 2003 as an incorporated administrative agency took over the NUPEC's projects for nuclear regulation and has been implementing FUBILA program that is for high burn up BWR full MOX cores. This paper presents an outline of the programs and a summary of the analysis results of the criticality of those experimental cores with major nuclear data libraries.

  4. Spent fuel handling and storage facility for an LWR fuel reprocessing plant

    Baker, W.H.; King, F.D.

    1979-01-01

    The facility will have the capability to handle spent fuel assemblies containing 10 MTHM/day, with 30% if the fuel received in legal weight truck (LWT) casks and the remaining fuel received in rail casks. The storage capacity will be about 30% of the annual throughput of the reprocessing plant. This size will provide space for a working inventory of about 50 days plant throughput and empty storage space to receive any fuel that might be in transit of the reprocessing plant should have an outage. Spent LWR fuel assemblies outside the confines of the shipping cask will be handled and stored underwater. To permit drainage, each water pool will be designed so that it can be isolated from the remaining pools. Pool water quality will be controlled by a filter-deionizer system. Radioactivity in the water will be maintained at less than or equal to 2 x 10 -4 Ci/m 3 ; conductivity will be maintained at 1 to 2 μmho/cm. The temperature of the pool water will be maintained at less than or equal to 40 0 C to retard algae growth and reduce evaporation. Decay heat will be transferred to the environment via a heat exchanger-cooling tower system

  5. Proceedings of the IAEA specialists` meeting on cracking in LWR RPV head penetrations

    Pugh, C.E.; Raney, S.J. [comps.] [Oak Ridge National Lab., TN (United States)

    1996-07-01

    This report contains 17 papers that were presented in four sessions at the IAEA Specialists` meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately.

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

    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)

  7. Accelerator driven light water fast reactor (revisiting to the accelerator LWR fuel regenerator)

    Takahashi, H.; Zhang, J.

    1999-01-01

    A tight-latticed, high-enriched Pu fuel reactor cooled by water or by super-critical steam has a high neutron economy, similar to that of Na-or Pb-cooled fast reactor. Operating in a subcritical condition by providing spallation neutrons, this Pu-fueled reactor can run safely, despite the positive coolant void coefficients. It can be used to transmute the proliferation-prone Pu into proliferation-resistive U-233 fuel using thorium as the fertile material. Rather than employing the large linear accelerator proposed for the LWR fuel regenerator studied in the INFCE program, a small circular accelerator, such as a cyclotron or a Fixed Field Alternating Gradient Synchrotron (FFAG), can run a large power reactor in a slightly subcritical reactor using control rods, on-line fuel reshuffling, and slightly graded proton-beam injection. Some thoughts on improving the reliability of the proton accelerator, on transmutation of the long-lived fission products of Tc-99, and I-129, and the future direction of the development of the fast reactor are discussed. (author)

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

    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

  9. Thermal-hydraulics technological strategy roadmap for LWR safety improvement and development

    Nakamura, Hideo; Arai, Kenji; Oikawa, Hirohide

    2015-01-01

    New version of the Thermal-Hydraulics Safety Evaluation Fundamental Technology Enhancement Strategy Roadmap (TH-RM) was developed by the Atomic Energy Society of Japan (AESJ) for LWR safety improvement and development. The 1st version of TH-RM was prepared in 2009 under collaboration of utilities, vendors, universities, research institutes and technical support organizations (TSO) for regulatory body. The revision was made by three sub-working groups (SWGs) by considering the lessons learned from the Fukushima Daiichi Accident. The 'safety assessment' SWG pursued development of computer codes for safety assessment. The 'fundamental technology' SWG pursued safety improvement and risk reduction via accident management (AM) measures by referring the technical map for severe accident (SA) established by the 'severe accident' SWG. Phenomena and components for counter-measures and/or proper prediction are identified by going through SA progression in both reactor and spent-fuel pool of PWR and BWR. Twelve important technology development subjects were identified, which include melt coolability enhancement to maintain integrity of containment vessel. Fact Sheet was developed to describe each of identified and selected R and D subjects. External hazards are also considered how to cope with from thermal-hydraulic safety point of view. This paper summarizes the revised TH-RM with several examples and future perspectives. (author)

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

    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)

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

    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)

  12. Deterministic estimation of crack growth rates in steels in LWR coolant environments

    Macdonald, D.D.; Lu, P.C.; Urquidi-Macdonald, M.

    1995-01-01

    In this paper, the authors extend the coupled environment fracture model (CEFM) for intergranular stress corrosion cracking (IGSCC) of sensitized Type 304SS in light water reactor heat transport circuits by incorporating steel corrosion, the oxidation of hydrogen, and the reduction of hydrogen peroxide, in addition to the reduction of oxygen (as in the original CEFM), as charge transfer reactions occurring on the external surfaces. Additionally, the authors have incorporated a theoretical approach for estimating the crack tip strain rate, and the authors have included a void nucleation model to account for ductile failure at very negative potentials. The key concept of the CEFM is that coupling between the internal and external environments, and the need to conserve charge, are the physical and mathematical constraints that determine the rate of crack advance. The model provides rational explanations for the effects of oxygen, hydrogen peroxide, hydrogen, conductivity, stress intensity, and flow velocity on the rate of crack growth in sensitized Type 304 in simulated LWR in-vessel environments. They propose that the CEFM can serve as the basis of a deterministic method for estimating component life times

  13. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

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

    2015-01-01

    In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial 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

  14. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

  15. Neutron dosimetry at nuclear power plants with light water reactors (LWR)

    Hofmann, B.; Schwarz, W.; Burgkhardt, B.; Piesch, E.

    1989-02-01

    During nuclear start-up of the Muelheim-Kaerlich nuclear power plant in 1986 the neutron radiation fields in the primary and auxiliary component rooms of the containment were investigated using the Single Sphere Albedo Technique and additional measurement techniques. For personnel monitoring albedo neutron dosemeters were used consisting of thermoluminescent detectors and track etch detectors combined with boron converters. Results: (1) The neutron radiation fields reach dose rate values up to 1000 mSv/h at the sleeves of the reactor coolant pipes, in the refuelling pool and the reactor cavity sump. The neutron component varies between 10% in the steam generator rooms up to 92% in the refuelling pool. (2) The mean value of the effective neutron energy at the different locations was found to be about 100 keV. Thermal neutrons contribute with about 10% to the area dose. (3) By direct intercomparisons and different evaluation methods of the Single Sphere Albedo Dosemeter it was shown, that rem-counters used within routine monitoring in the mixed radiation fields of the LWR overestimate the neutron dose rate only insignificantly (+20%) and are therefore usable for practical radiation protection work. (4) The sensitivity of albedo neutron dosemeters allows the detection of neutrons above 10 μSv. The contribution of neutrons to the total personnel dose was 25% in maximum. For the evaluation of albedo detectors a constant calibration factor can be applied. (orig./HP) [de

  16. Proceedings of the IAEA specialists' meeting on cracking in LWR RPV head penetrations

    Pugh, C.E.; Raney, S.J.

    1996-07-01

    This report contains 17 papers that were presented in four sessions at the IAEA Specialists' meeting on Cracking in LWR RPV Head Penetrations held at ASTM Headquarters in Philadelphia on May 2-3, 1995. The papers are compiled here in the order that presentations were made in the sessions, and they relate to operational observations, inspection techniques, analytical modeling, and regulatory control. The goal of the meeting was to allow international experts to review experience in the field of ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. The emphasis was to allow a better understanding of RPV material behavior, to provide guidance supporting reliability and adequate performance, and to assist in defining directions for further investigations. The international nature of the meeting is illustrated by the fact that papers were presented by researchers from 10 countries. There were technical experts present form other countries who participated in discussions of the results presented. This present document incorporates the final version of the papers as received from the authors. The final chapter includes conclusions and recommendations. Individual papers have been cataloged separately

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

    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)

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

    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

  19. Cross-section covariance propagation for LWR fuel cells in one and two dimensions - 308

    Ball, M.; Novog, D.R.; Parisi, C.; D'Auria, F.

    2010-01-01

    Within the framework of the Uncertainty Analysis in Modeling (UAM) for Design, Operation and Safety Analysis of LWRs Benchmark sponsored by the OECD/NEA, a tool has been developed for the propagation of covariance uncertainty through resonance self-shielding and other neutron kinetics calculations using a direct, cross-section generation and substitution approach. The motivation behind the work described in this paper was to develop a portable uncertainty propagation tool that could be easily implemented with several neutron kinetics codes, without relying on detailed knowledge of the internal workings of those codes or access to adjoint solutions. Implemented initially with the SCALE code package, 'self-shielded' covariance matrices for common LWR fuel cells have been calculated, as well as contributions to K eff uncertainty by selected neutron cross-sections and processes in both one and two dimensions. The one dimensional results generated by the tool are compared against those obtained using the TSUNAMI-1D module of SCALE in order to verify the efficacy of the methodology. One-dimensional results show good agreement with TSUNAMI-1D, but there is also an indication that the loss of dimensionality corresponding to one-dimensional equivalent geometries of two-dimensional fuel cells may lead to significant changes in the calculated uncertainty on K eff arising from particular neutron-nuclide reactions. (authors)

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

    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

  1. Consideration on the partial moderation in criticality safety analysis of LWR fresh fuel storage

    Tanaka, S.; Tanimoto, R.; Suzuki, K.; Ishitobi, M.

    1987-01-01

    In criticality safety analyses of fuel fabrication facilities, neutron effective multiplication factor (k eff ) of a storage vault has been calculated assuming ''partial moderation'' in whole space (hereafter reffered to as unlimited partial moderation). Where the enrichment of fuels to be stored is about 3.5 % or less, calculated k eff is usually low enough to show subcriticality even in unlimited partial moderation. However, it is scheduled to elevate LWR fuels enrichment for economical higher burnup and the unlimited partial moderation would require to introduce neutron absorbers to maintain subcriticality. It is clear that this causes economical disadvantages, and hence we reconsidered this assumption to avoid such a condition. Reconsideration of the unlimited partial moderation was carried out in following steps. (1) Water quantity to be assumed in atmosphere to obtain criticality was revealed too much to realize. (2) Typical realistic water quantity in atmosphere was estimated to apply as an alternative assumption. (3) A fresh fuel assembly storage was chosen as a model array and calculations with lattice code WIMS-D 1 and Monte Calro code KENO-IV 2 were performed to compare new alternative assumption with the unlimited one. As results of the above calculations, maximum k eff of the array under the new assumption was remarkably reduced to the value less than 0.95 though the maximum k eff under the unlimited one was higher than 1.0. (author)

  2. Poolside inspection, repair and reconstitution of LWR fuel elements. Proceedings of a Technical Committee meeting

    1998-11-01

    The Technical Committee Meeting on Poolside Inspection, repair and reconstruction of LWR Fuel Elements was organize by IAEA upon the recommendations of the International Working Group on Fuel performance Technology and held in Switzerland in October 1997. The purpose of the Meeting was to review the state of art in the area of poolside inspection, repair and reconstruction of light water fuel elements and to evaluate the progress achieved in this area since previous IAEA Meetings on the same topic in 1981 and 1984. The Meeting provided a forum on exchange of information between utilities, fuel designers and other authorities and specialists on a topic of current interest and real concern to industries in many Member States. The respective technologies are widely used or planned to be used in order to identify elementary major causes of fuel failure and to improve fuel utilization by repair and subsequent reuse of fuel elements. The Proceedings includes papers presented at the Meeting each described by a separate abstract

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

    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.

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

    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.

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

    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

  6. Incorporating Skew into RMS Surface Roughness Probability Distribution

    Stahl, Mark T.; Stahl, H. Philip.

    2013-01-01

    The standard treatment of RMS surface roughness data is the application of a Gaussian probability distribution. This handling of surface roughness ignores the skew present in the surface and overestimates the most probable RMS of the surface, the mode. Using experimental data we confirm the Gaussian distribution overestimates the mode and application of an asymmetric distribution provides a better fit. Implementing the proposed asymmetric distribution into the optical manufacturing process would reduce the polishing time required to meet surface roughness specifications.

  7. Investigation of surface roughness influence on hyperbolic metamaterial performance

    S. Kozik

    2014-12-01

    Full Text Available The main goal of this work was to introduce simple model of surface roughness which does not involve objects with complicated shapes and could help to reduce computational costs. We described and proved numerically that the influence of surface roughness at the interfaces in metal-dielectric composite materials could be described by proper selection of refractive index of dielectric layers. Our calculations show that this model works for roughness with RMS value about 1 nm and below.

  8. Constraining the roughness degree of slip heterogeneity

    Causse, Mathieu

    2010-05-07

    This article investigates different approaches for assessing the degree of roughness of the slip distribution of future earthquakes. First, we analyze a database of slip images extracted from a suite of 152 finite-source rupture models from 80 events (Mw = 4.1–8.9). This results in an empirical model defining the distribution of the slip spectrum corner wave numbers (kc) as a function of moment magnitude. To reduce the “epistemic” uncertainty, we select a single slip model per event and screen out poorly resolved models. The number of remaining models (30) is thus rather small. In addition, the robustness of the empirical model rests on a reliable estimation of kc by kinematic inversion methods. We address this issue by performing tests on synthetic data with a frequency domain inversion method. These tests reveal that due to smoothing constraints used to stabilize the inversion process, kc tends to be underestimated. We then develop an alternative approach: (1) we establish a proportionality relationship between kc and the peak ground acceleration (PGA), using a k−2 kinematic source model, and (2) we analyze the PGA distribution, which is believed to be better constrained than slip images. These two methods reveal that kc follows a lognormal distribution, with similar standard deviations for both methods.

  9. Spectrophotometric Examination of Rough Print Surfaces

    Erzsébet Novotny

    2011-05-01

    Full Text Available The objective was to assess the impact of the surface texture of individual creative paper types (coated or patternedon the quality of printing and to identify to what extent the various creative paper types require specific types ofspectrophotometers. We used stereomicroscopic images to illustrate unprinted and printed surfaces of creative papertypes. Surface roughness was measured to obtain data on the unevenness of surfaces. Spectrophotometric tests wereused to select the most suitable spectrophotometer from meters with different illumination setup for testing anygiven print. For the purpose of testing, we used spectrophotometers which are commonly available generally used totest print products for colour accuracy. With the improvement of measuring geometries, illumination setup, colourmeasurement becomes more and more capable of producing reliable results unaffected by surface textures. Our testshave proved this fact by showing that the GretagMacbeth Spectrolino with annular illumination is less sensitive tosurface texture than the X-Rite Spetrodensitometer and the Techkon SpetroDens with directional illumination. Furthertests have brought us to the conclusion that there is a difference even between the two devices with directionalillumination. While the X-Rite 530 Spectrodensitometer is more suitable for testing coated surfaces, the TechkonSpectroDens can come close to ΔE*ab values produced by the annular illuminated device for textured surfaces.

  10. Radiosensitivity of the rough-skinned newt

    Willis, D.L.; Lappenbusch, W.L.

    1974-01-01

    Newts were collected locally and maintained unfed at 10 0 C both before and after irradiation. Whole-body exposures ranging up to 80 kR were given with 250 kVp x rays, 300 kVp x rays, and 60 Co gamma rays. The mean survival time-exposure curve was sigmoid, following the typical appearance of such curves for mammalian species. A dose-independent region extending from roughly 700 to 5,000 R was evident. A variety of skin lesions and depigmentation effects were noted at exposures above 250 R, whose time of appearance decreased with increasing dose. A total of 832 newts were used in this phase of the study. Assuming that hematopoietic damage was primarily responsible for deaths below 700-800 R, additional studies of hematopoiesis were initiated. The effect of a 650 R-exposure on liver weight, spleen weight and concentration of circulating red blood cells for 1 1 / 2 months post-irradiation was assessed. The results indicate a progressive state of anemia. This was further substantiated by a study of 59 Fe incorporation by red blood cells at intervals up to one month for five exposures ranging up to 1000 R. (U.S.)

  11. Estimation of fracture roughness from the acoustic borehole televiewer image

    Bae, Dae Soek; Kim, Chun Soo; Kim, Kyung Soo; Park, Byung Yoon; Koh, Yong Kweon

    2000-12-01

    Estimation of fracture roughness - as one of the basic hydraulic fracture parameters - is very important in assessing ground water flow described by using discrete fracture network modeling. Former manual estimation of the roughness for each fracture surface of drill cores is above all a tedious, time-consuming work and will often cause some ambiguities of roughness interpretation partly due to the subjective judgements of observers, and partly due to the measuring procedure itself. However, recently, indebt to the highly reliable Televiewer data for the fracture discrimination, it has led to a guess to develop a relationship between the traditional roughness method based on a linear profiles and the method from the Televiewer image based on a ellipsoidal profile. Hence, the aim of this work is to develop an automatic evaluation algorithm for measuring the roughness from the Televiewer images. A highly reliable software named 'FRAFA' has been developed and realized to the extent that its utility merits. In the developing procedure, various problems - such as the examination of a new base line(ellipsoidal) for measuring the unevenness of fracture, the elimination of overlapping fracture signatures or noise, the wavelet estimation according to the type of fractures and the digitalization of roughness etc. - were considered. With these consideration in mind, the newly devised algorithm for the estimation of roughness curves showed a great potential not only for avoiding ambiguities of roughness interpretation but also for the judgement of roughness classification

  12. Estimating deep seafloor interface and volume roughness parameters using the multibeam-hydrosweep system

    Chakraborty, B.; Schenke, H.W.; Kodagali, V.N.; Hagen, R.

    composite roughness model, including water-sediment interface roughness and sediment volume roughness parameters the data was modeled. The model effectively uses the near normal incidence angle backscatter to determine the seafloor interface roughness...

  13. Fuzzy multi-project rough-cut capacity planning

    Masmoudi, Malek; Hans, Elias W.; Leus, Roel; Hait, Alain; Sotskov, Yuri N.; Werner, Frank

    2014-01-01

    This chapter studies the incorporation of uncertainty into multi-project rough-cut capacity planning. We use fuzzy sets to model uncertainties, adhering to the so-called possibilistic approach. We refer to the resulting proactive planning environment as Fuzzy Rough Cut Capacity Planning (FRCCP).

  14. Sub-Patch Roughness in Earthquake Rupture Investigations

    Zielke, Olaf

    2016-02-13

    Fault geometric complexities exhibit fractal characteristics over a wide range of spatial scales (<µm to >km) and strongly affect the rupture process at corresponding scales. Numerical rupture simulations provide a framework to quantitatively investigate the relationship between a fault\\'s roughness and its seismic characteristics. Fault discretization however introduces an artificial lower limit to roughness. Individual fault patches are planar and sub-patch roughnessroughness at spatial scales below fault-patch size– is not incorporated. Does negligence of sub-patch roughness measurably affect the outcome of earthquake rupture simulations? We approach this question with a numerical parameter space investigation and demonstrate that sub-patch roughness significantly modifies the slip-strain relationship –a fundamental aspect of dislocation theory. Faults with sub-patch roughness induce less strain than their planar-fault equivalents at distances beyond the length of a slipping fault. We further provide regression functions that characterize the stochastic effect sub-patch roughness.

  15. Heat transfer and pressure drop in microchannels with random roughness

    Pelevic, N.; van der Meer, Theodorus H.

    2016-01-01

    The effect of surface roughness on heat transfer and fluid flow phenomena within a microchannel has been investigated by using the lattice Boltzmann method. The surface roughness has been generated by using Gaussian function. Gaussian function is an efficient and convenient method to create surface

  16. Assessing of channel roughness and temperature variations on ...

    Assessing of channel roughness and temperature variations on wastewater quality parameters using numerical modeling. ... According to the obtained results, nitrate (NO3) has a decreasing trend when the Manning Roughness Coefficient (N) is higher than 0.04 along the channel, but is reduced when “N” is less than 0.04.

  17. Gliding Swifts Attain Laminar Flow over Rough Wings

    Lentink, D.; Kat, de R.

    2014-01-01

    Swifts are among the most aerodynamically refined gliding birds. However, the overlapping vanes and protruding shafts of their primary feathers make swift wings remarkably rough for their size. Wing roughness height is 1–2% of chord length on the upper surface—10,000 times rougher than sailplane

  18. Surface roughness effects on heat transfer in Couette flow

    Elia, G.G.

    1981-01-01

    A cell theory for viscous flow with rough surfaces is applied to two basic illustrative heat transfer problems which occur in Couette flow. Couette flow between one adiabatic surface and one isothermal surface exhibits roughness effects on the adiabatic wall temperature. Two types of rough cell adiabatic surfaces are studied: (1) perfectly insulating (the temperature gradient vanishes at the boundary of each cell); (2) average insulating (each cell may gain or lose heat but the total heat flow at the wall is zero). The results for the roughness on a surface in motion are postulated to occur because of fluid entrainment in the asperities on the moving surface. The symmetry of the roughness effects on thermal-viscous dissipation is discussed in detail. Explicit effects of the roughness on each surface, including combinations of roughness values, are presented to enable the case where the two surfaces may be from different materials to be studied. The fluid bulk temperature rise is also calculated for Couette flow with two ideal adiabatic surfaces. The effect of roughness on thermal-viscous dissipation concurs with the viscous hydrodynamic effect. The results are illustrated by an application to lubrication. (Auth.)

  19. A Meta-Analysis: Acoustic Measurement of Roughness and Breathiness

    v. Latoszek, Ben Barsties; Maryn, Youri; Gerrits, Ellen; De Bodt, Marc

    2018-01-01

    Purpose: Over the last 5 decades, many acoustic measures have been created to measure roughness and breathiness. The aim of this study is to present a meta-analysis of correlation coefficients (r) between auditory-perceptual judgment of roughness and breathiness and various acoustic measures in both sustained vowels and continuous speech. Method:…

  20. Road roughness evaluation using in-pavement strain sensors

    Zhang, Zhiming; Deng, Fodan; Huang, Ying; Bridgelall, Raj

    2015-11-01

    The international roughness index (IRI) is a characterization of road roughness or ride quality that transportation agencies most often report. The prevalent method of acquiring IRI data requires instrumented vehicles and technicians with specialized training to interpret the results. The extensive labor and high cost requirements associated with the existing approaches limit data collection to at most once per year for portions of the national highway system. Agencies characterize roughness only for some secondary roads but much less frequently, such as once every five years, resulting in outdated roughness information. This research developed a real-time roughness evaluation approach that links the output of durable in-pavement strain sensors to prevailing indices that summarize road roughness. Field experiments validated the high consistency of the approach by showing that it is within 3.3% of relative IRI estimates. After their installation and calibration during road construction, the ruggedized strain sensors will report road roughness continuously. Thus, the solution will provide agencies a real-time roughness monitoring solution over the remaining service life of road assets.