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Sample records for modern bwr cladding

  1. BWR fuel clad behaviour following LOCA

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R.

    1996-01-01

    Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. (author). 7 refs, 6 figs, 3 tabs

  2. A model for hydrogen pickup for BWR cladding materials

    International Nuclear Information System (INIS)

    Hede, G.; Kaiser, U.

    2001-01-01

    It has been observed that rod elongation is driven by the hydrogen pickup but not by corrosion as such. Based on this a non-destructive method to determine clad hydrogen concentration has been developed. The method is based on the observation that there are three different mechanisms behind the rod growth: the effect of neutron irradiation on the Zircaloy microstructure, the volume increase of the cladding as an effect of hydride precipitation and axial pellet-cladding-mechanical-interaction (PCMI). The derived correlation is based on the experience of older cladding materials, inspected at hot-cell laboratories, that obtained high hydrogen levels (above 500 ppm) at lower burnup (assembly burnup below 50 MWd/kgU). Now this experience can be applied, by interpolation, on more modern cladding materials with a burnup beyond 50 MWd/kgU by analysis of the rod growth database of the respective cladding materials. Hence, the method enables an interpolation rather than an extrapolation of present day hydrogen pickup database, which improves the reliability and accuracy. Further, one can get a good estimate of the hydrogen pickup during an ongoing outage based on a non-destructive method. Finally, rod growth measurements are normally performed for a large population of rods, hence giving a good statistics compared to examination of a few rods at a hot cell. (author)

  3. Fuel cladding tube and fuel rod for BWR type reactor

    International Nuclear Information System (INIS)

    Urata, Megumu; Mitani, Shinji.

    1995-01-01

    A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom (depth of the groove) in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube (depth of the groove) is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. (I.N.)

  4. Cladding tube of fuel rod for a BWR type reactor

    International Nuclear Information System (INIS)

    Nakayama, Hitoshi; Fujie, Kunio; Kuwahara, Heikichi; Hirai, Tadamasa; Kakizaki, Kimio.

    1976-01-01

    Object: To form a cladding tube wall with tunnels in communication with the exterior through a number of small-diameter openings to rapidly disperse a large quantity of heat thereby providing high density of the fuel rod. Structure: Tunnels adjacent to each other are provided under the skin in contact with cooling liquid of a cladding tube, and a number of openings through which said tunnels and the periphery of the cladding tube are placed in communication are formed, said openings each having its section smaller than that of said tunnel. With this arrangement, the cooling water entered the tunnel through some of small diameter openings absorbs heat of the fuel rod to be vaporized, which is flown out into the cooling water through the other small diameter openings and formed into vapor bubbles which move up for release of heat. (Taniai, N.)

  5. Improvement technique of sensitized HAZ by GTAW cladding applied to a BWR power plant

    International Nuclear Information System (INIS)

    Tujimura, Hiroshi; Tamai, Yasumasa; Furukawa, Hideyasu; Kurosawa, Kouichi; Chiba, Isao; Nomura, Keiichi.

    1995-01-01

    A SCC(Stress Corrosion Cracking)-resistant technique, in which the sleeve installed by expansion is melted by GTAW process without filler metal with outside water cooling, was developed. The technique was applied to ICM (In-Core Monitor) housings of a BWR power plant in 1993. The ICM housings of which materials are type 304 Stainless Steels are sensitized with high tensile residual stresses by welding to the RPV (Reactor Pressure Vessel). As the result, ICM housings have potential of SCC initiation. Therefore, the improvement technique resistant to SCC was needed. The technique can improve chemical composition of the housing inside and residual stresses of the housing outside at the same time. Sensitization of the housing inner surface area is eliminated by replacing low-carbon with proper-ferrite microstructure clad. High tensile residual stresses of housing outside surface area is improved into compressive side. Compressive stresses of outside surface are induced by thermal stresses which are caused by inside cladding with outside water cooling. The clad is required to be low-carbon metal with proper ferrite and not to have the new sensitized HAZ (Heat Affected Zone) on the surface by cladding. The effect of the technique was qualified by SCC test, chemical composition check, ferrite content measurement and residual stresses measurement etc. All equipment for remote application were developed and qualified, too. The technique was successfully applied to a BWR plant after sufficient training

  6. Effect of zinc injection on BWR fuel cladding corrosion. Pt. 1. Study on an accelerated corrosion condition to evaluate corrosion resistance of zircaloy-2 fuel cladding

    International Nuclear Information System (INIS)

    Kawamura, Hirotaka; Kanbe, Hiromu; Furuya, Masahiro

    2002-01-01

    Japanese BWR utilities have a plan to apply zinc injection to the primary coolant in order to reduce radioactivity accumulation on the structure. Prior to applying the zinc injection to BWR plants, it is necessary to evaluate the effect of zinc injection on corrosion resistance of fuel cladding. The objective of this report was to examine the accelerated corrosion condition for evaluation of BWR fuel cladding corrosion resistance under non-irradiated conditions, as the first step of a zinc injection evaluation study. A heat transfer corrosion test facility, in which a two phase flow condition could be achieved, was designed and constructed. The effects of heat flux, void fraction and solution temperature on BWR fuel cladding corrosion resistance were quantitatively investigated. The main findings were as follows. (1) In situ measurements using high speed camera and a void sensor together with one dimensional two phase flow analysis results showed that a two phase flow simulated BWR core condition can be obtained in the corrosion test facility. (2) The heat transfer corrosion test results showed that the thickness of the zirconium oxide layer increased with increasing solution temperature and was independent of heat flux and void fraction. The corrosion accelerating factor was about 2.5 times in the case of a temperature increase from 288degC to 350degC. (author)

  7. BWR stability: analysis of cladding temperature for high amplitude oscillations - 146

    International Nuclear Information System (INIS)

    Pohl, P.; Wehle, F.

    2010-01-01

    Power oscillations associated with density waves in boiling water reactors (BWRs) have been studied widely. Industrial research in this area is active since the invention of the first BWR. Stability measurements have been performed in various plants during commissioning phase but especially the magnitude and divergent nature of the oscillations during the LaSalle Unit 2 nuclear power plant event on March 9, 1988, renewed concern about the state of knowledge on BWR instabilities and possible consequences to fuel rod integrity. The objective of this paper is to present a simplified stability tool, applicable for stability analysis in the non-linear regime, which extends to high amplitude oscillations where inlet reverse flow occurs. In case of high amplitude oscillations a cyclical dryout and rewetting process at the fuel rod may take place, which leads in turn to rapid changes of the heat transfer from the fuel rod to the coolant. The application of this stability tool allows for a conservative determination of the fuel rod cladding temperature in case of high amplitude oscillations during the dryout / re-wet phase. Moreover, it reveals in good agreement to experimental findings the stabilizing effect of the reverse bundle inlet flow, which might be obtained for large oscillation amplitudes. (authors)

  8. Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

    International Nuclear Information System (INIS)

    Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

    1990-01-01

    Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

  9. Modern technology applied in the advanced BWR (ABWR)

    International Nuclear Information System (INIS)

    Hucik, S.A.

    1988-01-01

    The advanced boiling water reactor (ABWR) represents the next generation of light water reactors (LWR) to be introduced into commercial operation in the 1990's. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technology and improvements based on worldwide experience, and extensive design and test and development programs. This paper discusses how the ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. A technical evaluation of the ABWR shows its superiority in terms of performance characteristics and economics relative to current LWR designs

  10. Carbon 14 distribution in irradiated BWR fuel cladding and released carbon 14 after aqueous immersion of 6.5 years

    Energy Technology Data Exchange (ETDEWEB)

    Sakuragi, T. [Radioactive Waste Management Funding and Research Center, Tsukishima 1-15-7, Chuo City, Tokyo, 104-0052 (Japan); Yamashita, Y.; Akagi, M.; Takahashi, R. [TOSHIBA Corporation, Ukishima Cho 4-1, Kawasaki Ward, Kawasaki, 210-0862 (Japan)

    2016-07-01

    Spent fuel cladding which is highly activated and strongly contaminated is expected to be disposed of in an underground repository. A typical activation product in the activated metal waste is carbon 14 ({sup 14}C), which is mainly generated by the {sup 14}N(n,p){sup 14}C reaction and produces a significant exposure dose due to the large inventory, long half-life (5730 years), rapid release rate, and the speciation and consequent migration parameters. In the preliminary Japanese safety case, the release of radionuclides from the metal matrix is regarded as the corrosion-related congruent release, and the cladding oxide layer is regarded as a source of instant release fraction (IRF). In the present work, specific activity of {sup 14}C was measured using an irradiated BWR fuel cladding (Zircaloy-2, average rod burnup of 41.6 GWd/tU) which has an external oxide film having a thickness of 25.3 μm. The {sup 14}C specific activity of the base metal was 1.49*10{sup 4} Bq/g, which in the corresponding burnup is comparable to values in the existing literature, which were obtained from various irradiated claddings. Although the specific activity in oxide was 2.8 times the base metal activity due to the additive generation by the {sup 17}O(n,α){sup 14}C reaction, the {sup 14}C abundance in oxide was less than 10% of total inventory. A static leaching test using the cladding tube was carried out in an air-tight vessel filled with a deoxygenated dilute NaOH solution (pH of 12.5) at room temperature. After 6.5 years, {sup 14}C was found in each leachate fraction of gas phase and dissolved organics and inorganics, the total of which was less than 0.01% of the {sup 14}C inventory of the immersed cladding tube. A simple calculation based on the congruent release with Zircaloy corrosion has suggested that the 96.7% of released {sup 14}C was from the external oxide layer and 3.3% was from the base Zircaloy metal. However, both the {sup 14}C abundance and the low leaching rate

  11. Grid frequency control capability of a modern BWR

    International Nuclear Information System (INIS)

    Gemst, Paul van

    If a large generating units is tripped or if an intertie with a large amount of imported power is disconnected, the other units in the grid are firstly required to withstand the disturbance. The rapid restoration of the balance between energy production and consumption constitutes the second requirement. Tests performed on the Scandinavian grid have shown that the frequency minimum is reached about five seconds after a load rejection. The amplitude of the frequency drop depends on the load rejected and the grid power: 800 MWe in a 22,000 MWe grid yields 0.5 Hz drop. The grid frequency then returns to a value just under nominal in another 10 to 15 seconds. These figures agree well with calculations made by the Swedish State Power Board. The tests have also shown that the turbine generator should not be automatically and immediately disconnected after a reactor scram. This is in close agreement with and confirms traditional ASEA ATOM design policy and is in effect in all ASEA ATOM designed plants. The stored energy and residual power in the turbine and reactor cause the energy production to decrease with a time constant of 5 to 10 seconds. This, in turn, means that the minimum grid frequency is reached after more than 10 seconds and thus doubles the time available for other generating units to incease their power in order to compensate for the production loss and restore grid frequency. To minimize the frequency deviation and to improve network stability, a rapid increase in power production from those units that participate in grid frequency control may be necessary. The BWR has proved to be excellent for this purpose. A rapid increase of the electric power can be obtained by simultaneously decreasing the pressure, using some of energy stored in the reactor, and rapidly increasing recirculation flow

  12. Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R. [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Williams, T. [EGL Laufenburg (Switzerland); Helmersson, S. [Westinghouse Atom (Sweden)

    2001-03-01

    The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

  13. Experimental investigation of control absorber blade effects in a modern 10x10 BWR assembly

    International Nuclear Information System (INIS)

    Jatuff, F.; Grimm, P.; Murphy, M.; Luethi, A.; Seiler, R.; Joneja, O.; Meister, A.; Geemert, R. van; Brogli, R.; Chawla, R.; Williams, T.; Helmersson, S.

    2001-01-01

    The accurate estimation of reactor physics parameters related to the presence of cruciform absorber blades. In Boiling Water Reactors (BWR) is important for safety assessment, and for achieving a flexible operation during the cycle. Characteristics which are affected strongly include the power distribution for controlled core regions and its impact on linear heat generation rate margins, as well as the build-up of plutonium, and its influence on core excess reactivity and the reactivity worth of the shutdown system. PSI and the Swiss Nuclear Utilities (UAK) are conducting an experimental reactor physics programme related to modern Light Water Reactor (LWR) fuel assemblies, as employed in the Swiss nuclear power plants: the so-called. LWR-PROTEUS Phase I project. A significant part of this project has been devoted to the characterization of highly heterogeneous BWR fuel elements in the presence of absorber blades. The paper presents typical results for the performance of modern lattice codes in the estimation of controlled assembly reaction rate distributions, the sensitivity to the geometrical and material characterization, and a preliminary comparison of reflected-test-zone calculations with experimental reaction rate distributions measured in a Westinghouse SVEA-96+ assembly under full-density water moderation conditions in the presence of Westinghouse boron-carbide absorber blades. (author)

  14. Control chart analysis of data regarding 0.2% yield strength (YS) and percent total circumferential elongation (%TCE) for zircaloy clad tubes for PHWR and BWR fuels

    International Nuclear Information System (INIS)

    Yadav, M.B.; Singh, Hari; Vaidyanathan, S.; Sood, D.D.; Raghavan, S.V.; Bandyopadhyay, A.K.; Kulkarni, P.G.

    1992-01-01

    Zircaloy cladding tubes for PHWR and BWR fuels are manufactured and tested at Nuclear Fuel Complex (NFC), Hyderabad. Atomic Fuels Division is carrying out the quality assurance of the fuels on behalf of Nuclear Power Corporation (NPC). In this paper an attempt has been made to assess whether the quality of the clad tubes has met the requirements specified for the two mechanical properties of the tubes namely 0.2% yield strength and percent total circumferential elongation using control chart technique. For this purpose data for about 100 lots in each case were used. Process means and process standard deviations for these properties and the control limits for the corresponding control charts were estimated. The main findings are: (i) In case of PHWR tubes the production quality level with respect to 0.2% YS is higher, while that in case of %TCE is lower causing rejection of lots. On the other hand in the case of BWR tubes the production quality levels with respect to both the properties are higher than the required one. (ii) With respect to 0.2% YS, in case of BWR tubes a change in the pattern of distribution is detected beyond the lot serial no.47. However in case of PHWR tubes, though the data falls into two groups, no such pattern is seen. A modification in the acceptance/rejection criterion of the lot has been suggested. It is also pointed out that to have a correct picture of the total variation it is necessary to study the within tube variation. (author). 4 figs, 2 tabs

  15. Analysis of effects of pellet-cladding bonding on trapping of the released fission gases in high burnup KKL BWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Brankov, Vladimir [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Khvostov, Grigori; Mikityuk, Konstantin [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Pautz, Andreas [Laboratory for Reactor Physics and Systems Behaviour at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Swiss Federal Institute of Technology Lausanne (EPFL), Route Cantonale, 1015 Lausanne (Switzerland); Restani, Renato; Abolhassani, Sousan [Laboratory for Nuclear Materials at the Paul Scherrer Institute, 5232 Villigen-PSI (Switzerland); Ledergerber, Guido [Kernkraftwerk Leibstadt, 5325 Leibstadt (Switzerland); Wiesenack, Wolfgang [Institutt for Energiteknikk - OECD Halden Reactor Project, Os Allé 5, 1777 Halden (Norway)

    2016-08-15

    Highlights: • Explanation for the scatter in measured fission gas release in high-BU BWR fuel rods. • Partial fuel-clad bond layer formation in high-BU BWR fuel. • Hypothesis for fission gas trapping facilitated by the pellet-cladding bond layer. • Correlation between burnup asymmetry and the quantity of trapped fission gas. • Implications of the trapped FG in LOCA transient. - Abstract: The first part of the paper presents results of a numerical analysis of the fuel behavior during base irradiation in the Kernkraftwerk Leibstadt Boiling Water Reactor (KKL BWR) using EPRI’s FALCON code coupled to GRSW-A – an advanced model for fuel swelling and fission gas release. Post-irradiation examinations conducted at the Paul Scherrer Institute’s (PSI) hot laboratory gave evidence of a distinct circumferential non-uniformity of local burnup at pellet surfaces. For several fuel samples, intact pellet-cladding bonding areas on the high burnup sides of the pellets at high burnup above ∼70 MWd/kgU were observed. It is hypothesized that a part of the fission gases, which are expected to be released by those areas, can be trapped and do not reach the rod plenum. In this paper, a simple approach to modeling of fission gas trapping is employed which reveals a potential correlation between the position of the rod within the fuel assembly (and therefore the degree of circumferential burnup non-uniformity) and the degree of fission gas trapping. A model is suggested to correlate the amount of locally trapped gas with the integral of the local contact pressure and the degree of circumferential burnup non-uniformity. The model is calibrated with available measurements of FGR from rod puncturing at the level of the plenums. In future work, the hypothesis about the axial distribution of trapped fission gas will be extrapolated to the Loss-Of-Coolant Accident (LOCA) analysis as an attempt to explain the fission gas release observed in some samples fabricated from

  16. Studies of Corrosion of Cladding Materials in Simulated BWR-environment Using Impedance Measurements. Part I: Measurements in the Pre-transition Region

    International Nuclear Information System (INIS)

    Forsberg, Stefan; Ahlberg, Elisabet; Andersson, Ulf

    2004-09-01

    The corrosion of three Zircaloy 2 cladding materials, LK2, LK2+ and LK3, have been studied in-situ in an autoclave using electrochemical impedance spectroscopy. Measurements were performed in simulated BWR water at temperatures up to 288 deg C. The impedance spectra were successfully modelled using equivalent circuits. When the oxide grew thicker during the experiments, a change-over from one to two time constants was seen, showing that a layered structure was formed. Oxide thickness, oxide conductivity and effective donor density were evaluated from the impedance data. The calculated oxide thickness at the end of the experiments was consistent with the value obtained from SEM. It was shown that the difference in oxide growth rate between the investigated materials is small in the pre-transition region. The effective donor density, which is a measure of electronic conductivity, was found to be lower for the LK3 material compared to the other two materials

  17. Dynamics of a BWR with inclusion of boiling nonlinearity, clad temperature and void-dependent core power removal: Stability and bifurcation characteristics of advanced heavy water reactor (AHWR)

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Dinkar, E-mail: dinkar@iitk.ac.in [Nuclear Engineering and Technology Program, Indian Institute of Technology Kanpur, Kanpur 208 016 (India); Kalra, Manjeet Singh, E-mail: drmanjeet.singh@dituniversity.edu.in [DIT University, Dehradun 248 009 (India); Wahi, Pankaj, E-mail: wahi@iitk.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Kanpur, Kanpur 208 016 (India)

    2016-11-15

    Highlights: • Simplified models with inclusion of the clad temperature are considered. • Boiling nonlinearity and core power removal have been modeled. • Method of multiple time scales has been used for nonlinear analysis to get the nature and amplitude of oscillations. • Incorporation of modeling complexities enhances the stability of system. • We find that reactors with higher nominal power are more desirable from the point of view of global stability. - Abstract: We study the effect of including boiling nonlinearity, clad temperature and void-dependent power removal from the primary loop in the mathematical modeling of a boiling water reactor (BWR) on its dynamic characteristics. The advanced heavy water reactor (AHWR) is taken as a case study. Towards this end, we have analyzed two different simplified models with different handling of the clad temperature. Each of these models has the necessary modifications pertaining to boiling nonlinearity and power removal from the primary loop. These simplified models incorporate the neutronics and thermal–hydraulic coupling. The effect of successive changes in the modeling assumptions on the linear stability of the reactor has been studied and we find that incorporation of each of these complexities in the model increases the stable operating region of the reactor. Further, the method of multiple time scales (MMTS) is exploited to carry out the nonlinear analysis with a view to predict the bifurcation characteristics of the reactor. Both subcritical and supercritical Hopf bifurcations are present in each model depending on the choice of operating parameters. These analytical observations from MMTS have been verified against numerical simulations. A parametric study on the effect of changing the nominal reactor power on the regions in the parametric space of void coefficient of reactivity and fuel temperature coefficient of reactivity with sub- and super-critical Hopf bifurcations has been performed for all

  18. An analysis of instabilities of nuclear-coupled density-wave in BWR using modern frequency-domain control theory

    International Nuclear Information System (INIS)

    Zhao Yangping; Gao Huahun; Fu Longzhou

    1991-01-01

    A state-of-the-art multi-variable frequency-domain model has been developed for analysis of instabilities of nuclear-coupled density-wave in BWR core. The characteristic locus method is used for analysing the stability of BWR. A computer code-NUCTHIA has been derived. The model has been tested against the existing experimental data and compared with results of past single-variable analyses. By using the NUCTHIA code, the investigations of effects of main system parameters on BWW core stability have also been made. All the results are consistent with the experimental data

  19. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Science.gov (United States)

    González-Robles, E.; Serrano-Purroy, D.; Sureda, R.; Casas, I.; de Pablo, J.

    2015-10-01

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO2 spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAPc) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  20. Influence of partial blockage of a BWR bundle on heat transfer, cladding temperature, and quenching during bottom flooding or top spraying under simulated LOCA conditions

    International Nuclear Information System (INIS)

    Brand, B.; Gaul, H.P.; Sarkar, J.

    1982-01-01

    In a test facility with two parallel boiling water reactor fuel assemblies, experiments were carried out with top spray and bottom flooding, simulating loss-of-coolant accident (LOCA) conditions. The flow area restriction, caused by the ballooning of fuel rod cladding within one of the bundles, was provided by blockage plates, which had reductions of 37% in one case and in a second series 70% of the flow area. Test parameters were system pressure (1, 5, and 10 bars), spray (0.68 and 1.02 m 3 /h) and flooding rates (1.5,2, and 3.3 cm/s), power input (520 and 614 kW), and the initial cladding temperature (600 and 800 0 C at midplane) of the heaters. The test results showed no significant variations from those without blockage, except in the blocked region. An enhancement of heat transfer was observed in a close region downstream from the blockage in cases such as bottom flooding and top spray tests. The results will serve the purpose of code verification for reactor LOCA analysis

  1. Fuel Performance Calculations for FeCrAl Cladding in BWRs

    Energy Technology Data Exchange (ETDEWEB)

    George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

  2. Status update of the BWR cask simulator

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Durbin, Samuel G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    The performance of commercial nuclear spent fuel dry storage casks are typically evaluated through detailed numerical analysis of the system's thermal performance. These modeling efforts are performed by the vendor to demonstrate the performance and regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have moved the storage location from above ground to below ground and significantly increased the maximum thermal load allowed in a cask in part by increasing the canister helium pressure. Previous cask performance validation testing did not capture these parameters. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern dry casks. These modern cask designs utilize elevated helium pressure in the sealed canister or are intended for subsurface storage. The BWR cask simulator (BCS) has been designed in detail for both the above ground and below ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below ground storage configurations

  3. Initial Cladding Condition

    International Nuclear Information System (INIS)

    Siegmann, E.

    2000-01-01

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  4. Cladding creepdown model for FRAPCON-2

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.

    1985-02-01

    This report presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in both a pressurized water reactor (PWR) and a boiling water reactor (BWR). This model accounts for variations in zircaloy cladding heat treatment; cold worked and stress relieved material, typically used in a PWR, and fully recrystallized material, typically used in a BWR. The model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. This report also presents a comparison between cladding creep calculations by this model and corresponding measurements from the KWU/CE program, ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the model calculates cladding creep strains well. The analyses of non-fueled rods by FRAPCON-2 show that the cladding creepdown model was correctly incorporated. Also, analysis of a PWR rod test case shows that the FRAPCON-2 code can analyze pellet-cladding mechanical interaction caused by cladding creepdown and fuel swelling

  5. Structural cladding /clad structures

    DEFF Research Database (Denmark)

    Beim, Anne

    2012-01-01

    Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure in the pr......Structural Cladding /Clad Structures: Studies in Tectonic Building Practice A. Beim CINARK – Centre for Industrialized Architecture, Institute of Architectural Technology, The Royal Danish Academy of Fine Arts School of Architecture, Copenhagen, Denmark ABSTRACT: With point of departure...... to analyze, compare, and discuss how these various construction solutions point out strategies for development based on fundamentally different mindsets. The research questions address the following issues: How to learn from traditional construction principles: When do we see limitations of tectonic maneuver......, to ask for more restrictive building codes. As an example, in Denmark there are series of increasing demands in the current building legislations that are focused at enhancing the energy performance of buildings, which consequently foster rigid insulation standards and ask for improvement of air...

  6. BWR 90: The ABB advanced BWR design

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced fight water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and the total power generation costs have been low. In the development of BWR 90 specific changes were introduced to the reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher dim that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Thus, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The work is scheduled for completion in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with the 'evolutionary' design BWR 90+. The primary design goal is to develop the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is

  7. Dissolution experiments of commercial PWR (52 MWd/kgU) and BWR (53 MWd/kgU) spent nuclear fuel cladded segments in bicarbonate water under oxidizing conditions. Experimental determination of matrix and instant release fraction

    Energy Technology Data Exchange (ETDEWEB)

    González-Robles, E., E-mail: ernesto.gonzalez-robles@kit.edu [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Serrano-Purroy, D. [European Commission - EC, Joint Research Centre (JRC), Institute for Transuranium Elements - ITU, Postfach 2340, D-76125 Karlsruhe (Germany); Sureda, R. [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Casas, I. [Chemical Engineering Department, Universitat Politècnica de Catalunya, Av. Diagonal 647, 08028 Barcelona (Spain); Pablo, J. de [CTM Centre Tecnològic, Plaça de la Ciència 2, 08243 Manresa (Spain); Chemical Engineering Department, Universitat Politècnica de Catalunya, Av. Diagonal 647, 08028 Barcelona (Spain)

    2015-10-15

    The denominated instant release fraction (IRF) is considered in performance assessment (PA) exercises to govern the dose that could arise from the repository. A conservative definition of IRF comprises the total inventory of radionuclides located in the gap, fractures, and the grain boundaries and, if present, in the high burn-up structure (HBS). The values calculated from this theoretical approach correspond to an upper limit that likely does not correspond to what it will be expected to be instantaneously released in the real system. Trying to ascertain this IRF from an experimental point of view, static leaching experiments have been carried out with two commercial UO{sub 2} spent nuclear fuels (SNF): one from a pressurized water reactor (PWR), labelled PWR, with an average burn-up (BU) of 52 MWd/kgU and fission gas release (FGR) of 23.1%, and one from a boiling water reactor (BWR), labelled BWR, with an average BU of and 53 MWd/kgU and FGR of 3.9%. One sample of each SNF, consisting of fuel and cladding, has been leached in bicarbonate water during one year under oxidizing conditions at room temperature (25 ± 5)°C. The behaviour of the concentration measured in solution can be divided in two according to the release rate. All radionuclides presented an initial release rate that after some days levels down to a slower second one, which remains constant until the end of the experiment. Cumulative fraction of inventory in aqueous phase (FIAP{sub c}) values has been calculated. Results show faster release in the case of the PWR SNF. In both cases Np, Pu, Am, Cm, Y, Tc, La and Nd dissolve congruently with U, while dissolution of Zr, Ru and Rh is slower. Rb, Sr, Cs and Mo, dissolve faster than U. The IRF of Cs at 10 and 200 days has been calculated, being (3.10 ± 0.62) and (3.66 ± 0.73) for PWR fuel, and (0.35 ± 0.07) and (0.51 ± 0.10) for BWR fuel.

  8. High burnup (41 - 61 GWd/tU) BWR fuel behavior under reactivity initiated accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Takehiko; Kusagaya, Kazuyuki; Yoshinaga, Makio; Uetsuka, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-12-01

    High burnup boiling water reactor (BWR) fuel was pulse irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate fuel behavior under cold startup reactivity initiated accident (RIA) conditions. Temperature, deformation, failure, and fission gas release behavior under the simulated RIA condition was studied in the tests. Fuel failure due to pellet-cladding mechanical interaction (PCMI) did not occur in the tests with typical domestic BWR fuel at burnups up to 56 GWd/tU, because they had limited cladding embrittlement due to hydrogen absorption of about 100 ppm or less. However, the cladding failure occurred in tests with fuel at a burnup of 61 GWd/tU, in which the peak hydrogen content in the cladding was above 150 ppm. This type of failure was observed for the first time in BWR fuels. The cladding failure occurred at fuel enthalpies of 260 to 360 J/g (62 to 86 cal/g), which were higher than the PCMI failure thresholds decided by the Japanese Nuclear Safety Commission. From post-test examinations of the failed fuel, it was found that the crack in the BWR cladding progressed in a manner different from the one in PWR cladding failed in earlier tests, owing to its more randomly oriented hydride distribution. Because of these differences, the BWR fuel was judged to have failed at hydrogen contents lower than those of the PWR fuel. Comparison of the test results with code calculations revealed that the PCMI failure was caused by thermal expansion of pellets, rather than by the fission gas expansion in the pellets. The gas expansion, however, was found to cause large cladding hoop deformation later after the cladding temperature escalated. (author)

  9. Parallel channel effects under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Suzuki, H.; Hatamiya, S.; Murase, M.

    1988-01-01

    Due to parallel channel effects, different flow patterns such as liquid down-flow and gas up-flow appear simultaneously in fuel bundles of a BWR core during postulated LOCAs. Applying the parallel channel effects to the fuel bundle, water drain tubes with a restricted bottom end have been developed in order to mitigate counter-current flow limiting and to increase the falling water flow rate at the upper tie plate. The upper tie plate with water drain tubes is an especially effective means of increasing the safety margin of a reactor with narrow gaps between fuel rods and high steam velocity at the upper tie plate. The characteristics of the water drain tubes have been experimentally investigated using a small-scaled steam-water system simulating a BWR core. Then, their effect on the fuel cladding temperature was evaluated using the LOCA analysis program SAFER. (orig.)

  10. The BWR Stability Issue

    International Nuclear Information System (INIS)

    D'Auria, F.

    2008-01-01

    The purpose of this paper is to supply general information about Boiling Water Reactor (BWR) stability. The main concerned topics are: phenomenological aspects, experimental database, modelling features and capabilities, numerical models, three-dimensional modelling, BWR system performance during stability, stability monitoring and licensing aspects.

  11. Future possibilities of SUSEN technologies for R&D of nuclear fuel cladding

    International Nuclear Information System (INIS)

    Mikloš, M.

    2015-01-01

    R&D possibilities with nuclear fuel cladding were discussed in this paper. The availability of 10 MWT reactor with BWR and PWR loops having chemistry control was described. Activity transport and fuel cladding corrosion can be investigated in this facility including PIE. The facility has hot cells and the laboratory is expected to start in 2017

  12. Compact modular BWR (CM-BWR)

    International Nuclear Information System (INIS)

    Fennern, Larry; Boardman, Charles; Carroll, Douglas G.; Hida, Takahiko

    2003-01-01

    A preliminary assessment has shown that a small 350 MWe BWR reactor can be placed within a close fitting steel containment vessel that is 7.1 meters inside diameter. This allows the technology and manufacturing capability currently used to fabricate large ABWR reactor vessels to be used to provide a factory fabricated containment vessel for a 350 MWe BWR. When a close fitted steel containment is combined with a passive closed loop isolation condenser system and a natural circulating reactor system that contains a large water inventory, primary system leaks cannot uncover the core. This eliminates many of the safety systems needed in response to a LOCA that are common to large, conventional plant designs including. Emergency Core Flooding, Automatic Depressurization System, Active Residual Heat Removal, Safety Related Auxiliary Cooling, Safety Related Diesel Generators, Hydrogen Re-Combiners, Ex-vessel Core Retention and Cooling. By fabricating the containment in a factory and eliminating most of the conventional safety systems, the construction schedule is shortened and the capital cost reduced to levels that would not otherwise be possible for a relatively small modular BWR. This makes the CM-BWR a candidate for applications where smaller incremental power additions are desired relative to a large ALWR or where the local infrastructure is not able to accommodate a conventional ALWR plant rated at 1350 MWe or more. This paper presents a preliminary design description of a Compact Modular BWR (CM-BWR) whose design features dramatically reduce the size and cost of the reactor building and associated safety systems. (author)

  13. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type

    International Nuclear Information System (INIS)

    Lucatero, M.A.; Hernandez L, H.

    2003-01-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  14. A regression model for zircaloy cladding in-reactor creepdown: Database, development, and assessment

    International Nuclear Information System (INIS)

    Shah, V.N.; Tolli, J.E.; Lanning, D.

    1987-01-01

    The paper presents a cladding deformation model developed to analyze cladding creepdown during steady state operation in a PWR and a BWR. This model accounts for variation in the zircaloy cladding heat treatments - cold worked and stress relieved material typically used in a PWR and fully recrystallized material typically used in a BWR. This model calculates cladding creepdown as a function of hoop stress, fast neutron flux, exposure time, and temperature. The paper also presents a comparison between cladding creep calculations by the creepdown model and corresponding test results from the KWU/CE program. ORNL HOBBIE experiments, and EPRI/Westinghouse Engineering cooperative project. The comparisons show that the creepdown model calculates cladding creep strains reasonably well. (orig./HP)

  15. Power ramp tests of BWR-MOX fuels

    International Nuclear Information System (INIS)

    Asahi, K.; Oguma, M.; Higuchi, S.; Kamimua, K.; Shirai, Y.; Bodart, S.; Mertens, L.

    1996-01-01

    Power ramp test of BWR-MOX and UO 2 fuel rods base irradiated up to about 60 GWd/t in Dodewaard reactor have been conducted in BR2 reactor in the framework of the international DOMO programme. The MOX pellets were provided by BN (MIMAS process) and PNC (MH method). The MOX fuel rods with Zr-liner and non-liner cladding and the UO 2 fuel rods with Zr-liner cladding remained intact during the stepwise power ramp tests to about 600 W/cm, even at about 60 GWd/t

  16. BWR fuel performance under advanced water chemistry conditions – a delicate journey towards zero fuel failures – a review

    International Nuclear Information System (INIS)

    Hettiarachchi, S.

    2015-01-01

    Boiling Water Reactors (BWRs) have undergone a variety of chemistry evolutions over the past few decades as a result of the need to control stress corrosion cracking of reactor internals, radiation fields and personnel exposure. Some of the advanced chemistry changes include hydrogen addition, zinc addition, iron reduction using better filtration technologies, and more recently noble metal chemical addition to many of the modern day operating BWRs. These water chemistry evolutions have resulted in changes in the crud distribution on fuel cladding material, Co-60 levels and the Rod oxide thickness (ROXI) measurements using the conventional eddy current techniques. A limited number of Post-Irradiation Examinations (PIE) of fuel rods that exhibited elevated oxide thickness using eddy current techniques showed that the actual oxide thickness by metallography is much lower. The difference in these observations is attributed to the changing magnetic properties of the crud affecting the rod oxide thickness measurement by the eddy current technique. This paper will review and summarize the BWR fuel cladding performance under these advanced and improved water chemistry conditions and how these changes have affected the goal to reach zero fuel failures. The paper will also provide a brief summary of some of the results of hot cell PIE, results of crud composition evaluation, crud spallation, oxide thickness measurements, hydrogen content in the cladding and some fuel failure observations. (author) Key Words: Boiling Water Reactor, Fuel Performance, Hydrogen Addition, Zinc Addition, Noble Metal Chemical Addition, Zero Leakers

  17. Material operating behaviour of ABB BWR control rods

    International Nuclear Information System (INIS)

    Rebensdorff, B.; Bart, G.

    2000-01-01

    The BWR control rods made by ABB use boron carbide (B 4 C and hafnium as absorber material within a cladding of stainless steel. The general behaviour under operation has proven to be very good. ABB and many of their control rod customers have performed extensive inspection programs of control rod behaviour. However, due to changes in the material properties under fast and thermal neutron irradiation defects may occur in the control rods at high neutron fluences. Examinations of irradiated control rod materials have been performed in hot cell laboratories. The examinations have revealed the defect mechanism Irradiation Assisted Stress Corrosion Cracking (IASCC) to appear in the stainless steel cladding. For IASCC to occur three factors have to act simultaneously. Stress, material sensitization and an oxidising environment. Stress may be obtained from boron carbide swelling due to irradiation. Stainless steel may be sensitized to intergranular stress corrosion cracking under irradiation. Normally the reactor environment in a BWR is oxidising. The presentation focuses on findings from hot cell laboratory work on irradiated ABB BWR control rods and studies of irradiated control rod materials in the hot cells at PSI. Apart from physical, mechanical and microstructural examinations, isotope analyses were performed to describe the local isotopic burnup of boron. Consequences (such as possible B 4 C washout) of a under operation in a ABB BWR, after the occurrence of a crack is discussed based on neutron radiographic examinations of control rods operated with cracks. (author)

  18. IFPE/IFA-432, Fission Gas Release, Mechanical Interaction BWR Fuel Rods, Halden

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1996-01-01

    Description: It contains data from experiments that have been performed at the IFE/OECD Halden Reactor Project, available for use in fuel performance studies. It covers experiments on thermal performance, fission product release, clad properties and pellet clad mechanical interaction. It includes also experimental data relevant to high burn-up behaviour. IFA-432: Measurements of fuel temperature response, fission gas release and mechanical interaction on BWR-type fuel rods up to high burn-ups. The assembly featured several variations in rod design parameters, including fuel type, fuel/cladding gap size, fill gas composition (He and Xe) and fuel stability. It contained 6 BWR-type fuel rods with fuel centre thermocouples at two horizontal planes, rods were also equipped with pressure transducers and cladding extensometers. Only data from 6 rods are compiled here

  19. Calculation of the linear heat generation rates which violate the thermomechanical limit of plastic deformation of the fuel cladding in function of the burn up of a BWR fuel rod type; Calculo de las razones de generacion de calor lineal que violen el limite termomecanico de deformacion plastica de la camisa en funcion del quemado de una barra combustible tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lucatero, M.A.; Hernandez L, H. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: mal@nuclear.inin.mx

    2003-07-01

    The linear heat generation rates (LHGR) for a BWR type generic fuel rod, as function of the burnup that violate the thermomechanical limit of circumferential plastic deformation of the can (canning) in nominal operation in stationary state of the fuel rod are calculated. The evaluation of the LHGR in function of the burnt of the fuel, is carried out under the condition that the deformation values of the circumferential plastic deformation of the can exceeds in 0.1 the thermomechanical value operation limit of 1%. The results of the calculations are compared with the generation rates of linear operation heat in function of the burnt for this fuel rod type. The calculations are carried out with the FEMAXI-V and RODBURN codes. The results show that for exhibitions or burnt between 0 and 16,000 M Wd/tU a minimum margin of 160.8 W/cm exists among LHGR (439.6 W/cm) operation peak for the given fuel and maximum LHGR of the fuel (calculated) to reach 1.1% of circumferential plastic deformation of the can, for the peak factor of power of 1.40. For burnt of 20,000 MWd/tU and 60,000 MWd/tU exist a margin of 150.3 and 298.6 W/cm, respectively. (Author)

  20. Design criteria for confidence in the manufacture of BWR fuel rods

    International Nuclear Information System (INIS)

    Anantharaman, K.; Basu, S.; Anand, A.K.; Mehta, S.K.

    Based on the experience of fuel manufacture for BWR type reactors in India, the parameters which need stringent quality control, are discussed. The design specifications of the fuel rods as well as the cladding material and tubes are reported. The defect mechanisms to be taken into account and the fuel failure in reference to the variation of mechanical properties of the cladding are also described. (K.B.)

  1. EPRI BWR Water Chemistry Guidelines Revision

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.

    2014-01-01

    BWRVIP-190: BWR Water Chemistry Guidelines – 2008 Revision has been revised. The revision committee consisted of U.S. and non-U.S. utilities (members of the BWR Vessel and Internals Protection (BWRVIP) Mitigation Committee), reactor system manufacturers, fuel suppliers, and EPRI and industry experts. The revised document, BWRVIP-190 Revision 1, was completely reformatted into two volumes, with a simplified presentation of water chemistry control, diagnostic and good practice parameters in Volume 1 and the technical bases in Volume 2, to facilitate use. The revision was developed in parallel and in coordination with preparation of the Fuel Reliability Guidelines Revision 1: BWR Fuel Cladding Crud and Corrosion. Guidance is included for plants operating under normal water chemistry (NWC), moderate hydrogen water chemistry (HWC-M), and noble metal application (GE-Hitachi NobleChem™) plus hydrogen injection. Volume 1 includes significant changes to BWR feedwater and reactor water chemistry control parameters to provide increased assurance of intergranular stress corrosion cracking (IGSCC) mitigation of reactor materials and fuel reliability during all plant conditions, including cold shutdown (≤200°F (93°C)), startup/hot standby (>200°F (93°C) and ≤ 10%) and power operation (>10% power). Action Level values for chloride and sulfate have been tightened to minimize environmentally assisted cracking (EAC) of all wetted surfaces, including those not protected by hydrogen injection, with or without noble metals. Chemistry control guidance has been enhanced to minimize shutdown radiation fields by clarifying targets for depleted zinc oxide (DZO) injection while meeting requirements for fuel reliability. Improved tabular presentations of parameter values explicitly indicate levels at which actions are to be taken and required sampling frequencies. Volume 2 provides the technical bases for BWR water chemistry control for control of EAC, flow accelerated corrosion

  2. Analysis of coaxial laser micro cladding processing conditions

    OpenAIRE

    Tarasova, Tatiana Vasilievna; Gvozdeva, Galina Olegovna; Nowotny, Steffen; Ableyeva, Riana R.; Dolzhikova, Evgenia Yu

    2018-01-01

    The laser build-up cladding is a well-known technique for repair, coatings and additive manufacturing tasks. Modern equipment for the laser cladding enables material to be deposited with the lateral resolution of about 100 μm and to manufacture miniature precise parts. However, the micro cladding regimes are unknown. Determination of these regimes is an expensive task as a well-known relation between laser cladding parameters and melt pool dimensions are changing by technology micro-miniaturi...

  3. Critical cladding radius for hybrid cladding modes

    Science.gov (United States)

    Guyard, Romain; Leduc, Dominique; Lupi, Cyril; Lecieux, Yann

    2018-05-01

    In this article we explore some properties of the cladding modes guided by a step-index optical fiber. We show that the hybrid modes can be grouped by pairs and that it exists a critical cladding radius for which the modes of a pair share the same electromagnetic structure. We propose a robust method to determine the critical cladding radius and use it to perform a statistical study on the influence of the characteristics of the fiber on the critical cladding radius. Finally we show the importance of the critical cladding radius with respect to the coupling coefficient between the core mode and the cladding modes inside a long period grating.

  4. Modern

    Directory of Open Access Journals (Sweden)

    A.V. Bagrov

    2014-06-01

    Full Text Available The article gives an overview of the most important problems of modern meteoric astronomy and briefly describes ways and methods of their solutions. Particular attention is paid to the construction and arrangement of meteoric video cameras intended for registration of the meteoric phenomena as the main method of obtaining reliable and objective observational data on the basis of which the solution of the described tasks is possible.

  5. Development of advanced BWR

    International Nuclear Information System (INIS)

    Toyota, Masatoshi

    1982-01-01

    The Japanese technology and domestic production of BWR type nuclear power plants have been established through the experiences in the construction and operation of BWRs in addition to the technical agreement with the General Electric Co. In early days, the plants experienced some trouble such as stress corrosion cracking and some inconvenience in the operation and maintenance. The government, electric power companies and BWR manufacturers have endeavored to standardize and improve the design of LWRs for the purpose of improving the safety, reliability and the rate of operation and reducing the radiation exposure dose of plant workers. The first and second stages of the standardization and improvement of LWRs have been completed. Five manufacturers of BWRs in the world have continued the conceptual design of a new version of BWR power plants. It was concluded that this is the most desirable version of BWR nuclear power stations, but the technical and economic evaluation must be made before the commercial application. Six electric power companies and three manufacturers of BWRs in Japan set up the organization to develop the technology in cooperation. The internal pump system, the new control rod drive mechanism and others are the main features. (Kako, I.)

  6. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

    1992-01-01

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  7. Fuel cladding behavior under rapid loading conditions

    Science.gov (United States)

    Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L.

    2016-02-01

    A modified burst test (MBT) was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident (RIA). A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile (BTD) transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor (PWR) cladding at both 296 and 553 K.

  8. Pie technique of LWR fuel cladding fracture toughness test

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  9. BWR stability analysis

    International Nuclear Information System (INIS)

    Valtonen, K.

    1990-01-01

    The objective of this study has been to examine TVO-I oscillation incident, which occured in February 22.1987 and to find out safety implications of oscillations in ATWS incidents. Calculations have been performed with RAMONA-3B and TRAB codes. RAMONA-3B is a BWR transient analysis code with three-dimencional neutron kinetics and nonequilibrium, nonhomogeneous thermal hydraulics. TRAB code is a one-dimencional BWR transient code which uses methods similar to RAMONA-3B. The results have shown that both codes are capable of analyzing of the oscillation incidents. Both out-of-phase and in-phase oscillations are possible. If the reactor scram fails (ATWS) during oscillations the severe fuel failures are always possible and the reactor core may exceed the prompt criticality

  10. TRAC-BWR development

    International Nuclear Information System (INIS)

    Weaver, W.L.; Rouhani, S.Z.

    1983-01-01

    The TRAC-BD1/MOD1 code containing many new or improved models has been assembled and is undergoing developmental assessment and testing and should be available shortly. The preparation of the manual for this code version is underway and should be available to the USNRC and their designated contractors by April of 1984. Finally work is currently underway on a fast running version of TRAC-BWR which will contain a one-dimensional neutron kinetics model

  11. Study on thermal performance and margins of BWR fuel elements

    International Nuclear Information System (INIS)

    Stosic, Zoran

    1999-01-01

    This paper contributes to developing a methodology of predicting and analyzing thermal performance and margins of Boiling Water Reactor (BWR) fuel assemblies under conditions of reaching high quality Boiling Crisis and subsequent post-dryout thermal hydraulics causing temperature excursion of fuel cladding. Operational margins against dryout and potential for increasing fuel performance with appropriate benefits are discussed. The philosophy of modeling with its special topics are demonstrated on the HECHAN (HEated CHannel ANalyzer) model as the state-of-art for thermal-hydraulics analysis of BWR fuel assemblies in pre- and post-dryout two-phase flow regimes. The scope of further work either being or has to be performed concerning implementation of new physical aspects, including domain extension of HECHAN model applications to the Pressurized Water Reactors (PWRs), is discussed. Finally, a comprehensive overview of the literature dealing with development of the model is given. (author)

  12. BWR AXIAL PROFILE

    International Nuclear Information System (INIS)

    Huffer, J.

    2004-01-01

    The purpose of this calculation is to develop axial profiles for estimating the axial variation in burnup of a boiling water reactor (BWR) assembly spent nuclear fuel (SNF) given the average burnup of an assembly. A discharged fuel assembly typically exhibits higher burnup in the center and lower burnup at the ends of the assembly. Criticality safety analyses taking credit for SNF burnup must account for axially varying burnup relative to calculations based on uniformly distributed assembly average burnup due to the under-burned tips. Thus, accounting for axially varying burnup in criticality analyses is also referred to as accounting for the ''end effect'' reactivity. The magnitude of the reactivity change due to ''end effect'' is dependent on the initial assembly enrichment, the assembly average burnup, and the particular axial profile characterizing the burnup distribution. The set of bounding axial profiles should incorporate multiple BWR core designs and provide statistical confidence (95 percent confidence that 95 percent of the population is bound by the profile) that end nodes are conservatively represented. The profiles should also conserve the overall burnup of the fuel assembly. More background on BWR axial profiles is provided in Attachment I

  13. Stone cladding engineering

    CERN Document Server

    Sousa Camposinhos, Rui de

    2014-01-01

    This volume presents new methodologies for the design of dimension stone based on the concepts of structural design while preserving the excellence of stonemasonry practice in façade engineering. Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements. Based on the Load and Resistance Factor Design Format (LRDF), minimum slab thickness formulae are presented that take into consideration stress concentrations analysis based on the Finite Element Method (FEM) for the most commonly used modern anchorage systems. Calculation examples allow designers to solve several anchorage engineering problems in a detailed and objective manner, underlining the key parameters. The design of the anchorage metal parts, either in stainless steel or aluminum, is also presented.

  14. Characteristics of axial splits in failed BWR fuel rods

    International Nuclear Information System (INIS)

    Lysell, G.; Grigoriev, V.

    2000-01-01

    Secondary cladding defects in BWR fuel sometimes have the shape of long axial cracks or ''splits''. Due to the large open UO 2 surfaces exposed to the water, fission product and UO 2 release to the coolant can reach excessive levels leading to forced shut downs to remove the failed fuel rods. A number of such fuel rods have been examined in Studsvik over the last 10 years. The paper describes observations from the PIE of long cracks and discusses the driving force of the cracks. Details such as starting cracks, macroscopic and microscopic fracture surface appearance, cross sections of cracks, hydride precipitates, location and degree of plastic deformation are given. (author)

  15. Repair and preventive maintenance technology for BWR reactor internals and piping

    International Nuclear Information System (INIS)

    Ootsubo, Tooru; Itou, Takashi; Sakashita, Akihiro

    2009-01-01

    Stress corrosion cracking of welding portion has found in many domestic and foreign BWR reactor internals and Primary Loop Recirculation piping. Also, repair and preventive maintenance technologies for SCC has been developed and/or adopted to BWRs in recent years. This paper introduces the sample of these technologies, such as seal-welding for SCC on BWR reactor internals, preventive maintenance technology for PLR piping such as Corrosion Resistant Cladding, Internal Polishing and Induction Heating Stress Improvement. These technologies are introduced on 'E-Journal of Advanced Maintenance', which is an international journal on a exclusive website of Japan Society of Maintenology. (author)

  16. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  17. BWR condensate filtration studies

    International Nuclear Information System (INIS)

    Wilson, J.A.; Pasricha, A.; Rekart, T.E.

    1993-09-01

    Poor removal of particulate corrosion products (especially iron) from condensate is one of the major problems in BWR systems. The presence of activated corrosion products creates ''hot spots'' and increases piping dose rates. Also, fuel efficiency is reduced and the risk of fuel failure is increased by the deposit of corrosion products on the fuel. Because of these concerns, current EPRI guidelines call for a maximum of 2 ppb of iron in the reactor feedwater with a level of 0.5 ppb being especially desirable. It has become clear that conventional deep bed resins are incapable of meeting these levels. While installation of prefilter systems is an option, it would be more economical for plants with naked deep beds to find an improved bead resin for use in existing systems. BWR condensate filtration technologies are being tested on a condensate side stream at Hope Creek Nuclear Generating Station. After two years of testing, hollow fiber filters (HFF) and fiber matrix filters (FMF), and low crosslink cation resin, all provide acceptable results. The results are presented for pressure drop, filtration efficiency, and water quality measurements. The costs are compared for backwashable non-precoat HFF and FMF. Results are also presented for full deep bed vessel tests of the low crosslink cation resin

  18. BWR type nuclear reactors

    International Nuclear Information System (INIS)

    Yamamoto, Toru.

    1987-01-01

    Purpose: To obtain reactor core characteristics with less changes in the excess reactivity due to fuel burnup even when the operation period varies. Constitution: In a BWR type reactor where fuel assemblies containing fuel rods incorporated with burnable poisons are arranged, the fuel assemblies are grouped into first fuel assemblies and second fuel assemblies. Then, the number of fuel rods incorporated with burnable poisons within the first fuel assemblies is made greater than that of the second fuel rods, while the concentration of the burnable poisons in the fuel rods incorporated with the burnable poisons in the first fuel assemblies is made lower than that of the fuel rods incorporated with the burnable poisons in the second fuel assemblies. In the BWR type reactor constituted in this way, the reactor core characteristics can be improved by changing the ratio between the first fuel assemblies and the second fuel assemblies charged to the reactor core, thereby decreasing the changes in the burnup of the excess reactivity. (Kamimura, M.)

  19. BWR zinc addition Sourcebook

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Alfred J.

    2014-01-01

    Boiling Water Reactors (BWRs) have been injecting zinc into the primary coolant via the reactor feedwater system for over 25 years for the purpose of controlling primary system radiation fields. The BWR zinc injection process has evolved since the initial application at the Hope Creek Nuclear Station in 1986. Key transitions were from the original natural zinc oxide (NZO) to depleted zinc oxide (DZO), and from active zinc injection of a powdered zinc oxide slurry (pumped systems) to passive injection systems (zinc pellet beds). Zinc addition has continued through various chemistry regimes changes, from normal water chemistry (NWC) to hydrogen water chemistry (HWC) and HWC with noble metals (NobleChem™) for mitigation of intergranular stress corrosion cracking (IGSCC) of reactor internals and primary system piping. While past reports published by the Electric Power Research Institute (EPRI) document specific industry experience related to these topics, the Zinc Sourcebook was prepared to consolidate all of the experience gained over the past 25 years. The Zinc Sourcebook will benefit experienced BWR Chemistry, Operations, Radiation Protection and Engineering personnel as well as new people entering the nuclear power industry. While all North American BWRs implement feedwater zinc injection, a number of other BWRs do not inject zinc. This Sourcebook will also be a valuable resource to plants considering the benefits of zinc addition process implementation, and to gain insights on industry experience related to zinc process control and best practices. This paper presents some of the highlights from the Sourcebook. (author)

  20. LBB application in Swedish BWR design

    Energy Technology Data Exchange (ETDEWEB)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P. [ABB Atom, Vaesteras (Sweden)

    1997-04-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions.

  1. LBB application in Swedish BWR design

    International Nuclear Information System (INIS)

    Kornfeldt, H.; Bjoerk, K.O.; Ekstroem, P.

    1997-01-01

    The protection against dynamic effects in connection with potential pipe breaks has been implemented in different ways in the development of BWR reactor designs. First-generation plant designs reflect code requirements in effect at that time which means that no piping restraint systems were designed and built into those plants. Modern designs have, in contrast, implemented full protection against damage in connection with postulated pipe breaks, as required in current codes and regulations. Moderns standards and current regulatory demands can be met for the older plants by backfitting pipe whip restraint hardware. This could lead to several practical difficulties as these installations were not anticipated in the original plant design and layout. Meeting the new demands by analysis would in this situation have great advantages. Application of leak-before-break criteria gives an alternative opportunity of meeting modem standards in reactor safety design. Analysis takes into account data specific to BWR primary system operation, actual pipe material properties, piping loads and leak detection capability. Special attention must be given to ensure that the data used reflects actual plant conditions

  2. Study of behavior on bonding and failure mode of pressurized and doped BWR fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1992-03-01

    The study of transient behavior on the bonding and the failure mode was made using the pressurized/doped 8 x 8 BWR type fuel rod. The dopant was mullite minerals consisted mainly of silicon and aluminum up to 1.5 w/o. Pressurization of the fuel rod with pure helium was made to the magnitude about 0.6 MPa. As a reference, the non-pressurized/non-doped 8 x 8 BWR fuel rod and the pressurized/7 x 7 BWR fuel rod up to 0.6 MPa were prepared. Magnitude of energy deposition given to the tested fuel rods was 248, 253, and 269 cal/g·fuel, respectively. Obtained results from the pulse irradiation in NSRR are as follows. (1) It was found from the experiment that alternation of the fuel design by the adoption of pressurization up to 0.6 MPa and the use of wider gap up to 0.38 mm could avoid the dopant BWR fuel from the overall bonding. The failure mode of the present dopant fuel was revealed to be the melt combined with rupture. (2) The time of fuel failure of the pressurized/doped 8 x 8 BWR fuel defected by the melt/rupture mode is of order of two times shorter than that of the pressurized/ 7 x 7 BWR defected by the rupture mode. Failure threshold of the pressurized/doped 8 x 8 BWR BWR tended to be lower than that of non-pressurized/non-doped 8 x 8 BWR one. Cracked area of the pressurized/doped 8 x 8 BWR was more wider and magnitude of oxidation at the place is relatively larger than the other tested fuels. (3) Failure mode of the non-pressurized/ 8 x 8 BWR fuel rod was the melt/brittle accompanied with a significant bonding at failed location. While, failure mode of the pressurized/ 7 x 7 BWR fuel rod was the cladding rupture accompanied with a large ballooning. No bonding at failed location of the latter was observed. (author)

  3. BWR 90 and BWR 90+: Two advanced BWR design generations from ABB

    International Nuclear Information System (INIS)

    Haukeland, S.; Ivung, B.; Pedersen, T.

    1999-01-01

    ABB has two evolutionary advanced light water reactors available today - the BWR 90 boiling water reactor and the System 80+ pressurised water reactor. The BWR 90 is based on the design, construction, commissioning and operation of the BWR 75 plants. The operation experience of the six plants of this advanced design has been very good. The average annual energy availability is above 90%, and total power generation costs have been low. When developing the BWR 90 specific changes were introduced to a reference design, to adapt to technological progress, new safety requirements and to achieve cost savings. The thermal power rating of BWR 90 is 3800 MWth (providing a nominal 1374 MWe net), slightly higher than that of the reference plant ABB Atom has taken advantage of margins gained using a new generation of its SVEA fuel to attain this power rating without major design modifications. The BWR 90 design was completed and offered to the TVO utility in Finland in 1991, as one of the contenders for the fifth Finnish nuclear power plant project. Hence, the design is available today for deployment in new plant projects. Utility views were incorporated through co-operation with the Finnish utility TVO, owner and operator of the two Olkiluoto plants of BWR 75 design. A review against the European Utility Requirement (EUR) set of requirements has been performed, since the design, in 1997, was selected by the EUR Steering Committee to be the first BWR to be evaluated against the EUR documents. The review work was completed in 1998. It will be the subject of an 'EUR Volume 3 Subset for BWR 90' document. ABB is continuing its BWR development work with an 'evolutionary' design called BWR 90+, which aims at developing the BWR as a competitive option for the anticipated revival of the market for new nuclear plants beyond the turn of the century, as well as feeding ideas and inputs to the continuous modernisation efforts at operating plants. The development is performed by ABB Atom

  4. Probabilistic assessment of spent-fuel cladding breach

    International Nuclear Information System (INIS)

    Foadian, H.; Rashid, Y.R.; Seager, K.D.

    1991-01-01

    A methodology for determining the probability spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR (GE 7 x 7) and a PWR (B ampersand W 15 x 15) assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire

  5. Probabilistic assessment of spent-fuel cladding breach

    International Nuclear Information System (INIS)

    Foadian, H.; Rashid, Y.R.; Seager, K.D.

    1992-01-01

    In this paper a methodology for determining the probability of spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR (GE 7 x 7) and a PWR (B and W 15 x 15) assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire

  6. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  7. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  8. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  9. Safety analysis of thorium-based fuels in the General Electric Standard BWR

    International Nuclear Information System (INIS)

    Colby, M.J.; Townsend, D.B.; Kunz, C.L.

    1980-06-01

    A denatured (U-233/Th)O 2 fuel assembly has been designed which is energy equivalent to and hardware interchangeable with a modern boiling water reactor (BWR) reference reload assembly. Relative to the reference UO 2 fuel, the thorium fuel design shows better performance during normal and transient reactor operation for the BWR/6 product line and will meet or exceed current safety and licensing criteria. Power distributions are flattened and thermal operating margins are increased by reduced steam void reactivity coefficients caused by U-233. However, a (U-233/Th)O 2 -fueled BWR will likely have reduced operating flexibility. A (U-235/Th)O 2 -fueled BWR should perform similar to a UO 2 -fueled BWR under all operating conditions. A (Pu/Th)O 2 -fueled BWR may have reduced thermal margins and similar accident response and be less stable than a UO 2 -fueled BWR. The assessment is based on comparisions of point model and infinite lattice predictions of various nuclear reactivity parameters, including void reactivity coefficients, Doppler reactivity coefficients, and control blade worths

  10. BWR type reactors

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1986-01-01

    Purpose: To enable to remove water not by way of mechanical operation in a reactor core and improve the fuel economy in BWR type reactors. Constitution: A hollow water removing rod of a cross-like profile made of material having a smaller neutron absorption cross section than the moderator is disposed to the water gap for each of unit structures composed of four fuel assemblies, and water is charged and discharged to and from the water removing rod. Water is removed from the water removing rod to decrease the moderators in the water gap to carry out neutron spectrum shift operation from the initial to the medium stage of reactor core cycles. At the final stage of the cycle, airs in the water removing rod are extracted and the moderator is introduced. The moderator is filled and the criticality is maintained with the accumulated nuclear fission materials. The neutron spectrum shift operation can be attained by eliminating hydrothermodynamic instability and using a water removing rod of a simple structure. (Horiuchi, T.)

  11. BWR emergency procedure guidelines

    International Nuclear Information System (INIS)

    Post, J.S.; Karner, E.F.; Stratman, R.A.

    1984-01-01

    This chapter describes plans for dealing with reactor accidents developed by the Boiling Water Reactor (BWR) Owners' Group in response to post-Three Mile Island US NRC requirements. The devised Emergency Procedure Guidelines (EPGs), applicable to all BWRs, are symptom-based rather than event-based. According to the EPGs, the operator does not need to identify what event is occurring in the plant in order to decide what action to take, but need only observe the symptoms (values and trends of key control parameters) which exist and take appropriate action to control these symptoms. The original objective was to provide reactor operator guidance in responding to a small break loss-of-coolant accident (LOCA), but subsequent revisions have included other types of reactor accidents. Topics considered include the reactor pressure vessel (RPV) control guideline, the primary containment control guideline, the secondary containment control guideline, the radioactivity release control guideline, multiple failures vs. the design basis, safe limits vs. technical specifications, the technical status, licensing, and implementation. The EPGs are based upon maintaining both adequate core cooling and primary containment integrity

  12. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  13. BWR type reactor

    International Nuclear Information System (INIS)

    Okano, Shigeru.

    1992-01-01

    In a BWR type reactor, control rod drives are disposed in the upper portion of a reactor pressure vessel, and a control rod guide tube is disposed in adjacent with a gas/liquid separator at a same height, as well as a steam separator is disposed in the control rod guide tube. The length of a connection rod can be shortened by so much as the control rod guide tube and the gas/liquid separator overlapping with each other. Since the control rod guide tube and the gas/liquid separator are at the same height, the number of the gas/liquid separators to be disposed is decreased and, accordingly, even if the steam separation performance by the gas/liquid separator is lowered, it can be compensated by the steam separator of the control rod guide tube. In view of the above, since the direction of emergent insertion of the control rod is not against gravitational force but it is downward direction utilizing the gravitational force, reliability for the emergent insertion of the control rod can be further improved. Further, the length of the connection rod can be minimized, thereby enabling to lower the height of the reactor pressure vessel. The construction cost for the nuclear power plant can be reduced. (N.H.)

  14. High Fidelity BWR Fuel Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Su Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    This report describes the Consortium for Advanced Simulation of Light Water Reactors (CASL) work conducted for completion of the Thermal Hydraulics Methods (THM) Level 3 milestone THM.CFD.P13.03: High Fidelity BWR Fuel Simulation. High fidelity computational fluid dynamics (CFD) simulation for Boiling Water Reactor (BWR) was conducted to investigate the applicability and robustness performance of BWR closures. As a preliminary study, a CFD model with simplified Ferrule spacer grid geometry of NUPEC BWR Full-size Fine-mesh Bundle Test (BFBT) benchmark has been implemented. Performance of multiphase segregated solver with baseline boiling closures has been evaluated. Although the mean values of void fraction and exit quality of CFD result for BFBT case 4101-61 agreed with experimental data, the local void distribution was not predicted accurately. The mesh quality was one of the critical factors to obtain converged result. The stability and robustness of the simulation was mainly affected by the mesh quality, combination of BWR closure models. In addition, the CFD modeling of fully-detailed spacer grid geometry with mixing vane is necessary for improving the accuracy of CFD simulation.

  15. First interim examination of defected BWR and PWR rods tested in unlimited air at 2290C

    International Nuclear Information System (INIS)

    Einziger, R.E.; Cook, J.A.

    1983-01-01

    A five-year whole rod test was initiated to investigate the long-term stability of spent fuel rods under a variety of possible dry storage conditions. Both PWR and BWR rods were included in the test. The first interim examination was conducted after three months of testing to determine if there was any degradation in those defected rods stored in an unlimited air atmosphere. Visual observations, diametral measurements and radiographic smears were used to assess the degree of cladding deformation and particulate dispersal. The PWR rod showed no measurable change from the pre-test condition. The two original artificial defects had not changed in appearance and there was no diametral growth of the cladding. One of the defects in BWR rod showed significant deformation. There was approximately 10% cladding strain at the defect site and a small axial crack had formed. The fuel in the defect did not appear to be friable. The second defect showed no visible change and no cladding strain. Following examination, the test was continued at 230 0 C. Another interim examination is planned during the summer of 1983. This paper discusses the details and meaning of the data from the first interim examination

  16. Correlation of waterside corrosion and cladding microstructure in high-burnup fuel and gadolinia rods

    International Nuclear Information System (INIS)

    Chung, H.M.

    1989-09-01

    Waterside corrosion of the Zircaloy cladding has been examined in high-burnup fuel rods from several BWRs and PWRs, as well as in 3 wt % gadolinia burnable poison rods obtained from a BWR. The corrosion behavior of the high-burnup rods was then correlated with results from a microstructural characterization of the cladding by optical, scanning-electron, and transmission-electron microscopy (OM, SEM, and TEM). OM and SEM examination of the BWR fuel cladding showed both uniform and nodular oxide layers 2 to 45 μm in thickness after burnups of 11 to 30 MWd/kgU. For one of the BWRs, which was operated at 307 degree C rather than the normal 288 degree C, a relatively thick (50 to 70 μm) uniform oxide, rather than nodular oxides, was observed after a burnup of 27 to 30 MWd/kgU. TEM characterization revealed a number of microstructural features that occurred in association with the intermetallic precipitates in the cladding metal, apparently as a result of irradiation-induced or -enhanced processes. The BWR rods that exhibited white nodular oxides contained large precipitates (300 to 700 nm in size) that were partially amorphized during service, indicating that a distribution of the large intermetallic precipitates is conductive to nodular oxidation. 23 refs., 9 figs

  17. Laser cladding with powder

    NARCIS (Netherlands)

    Schneider, M.F.; Schneider, Marcel Fredrik

    1998-01-01

    This thesis is directed to laser cladding with powder and a CO2 laser as heat source. The laser beam intensity profile turned out to be an important pa6 Summary rameter in laser cladding. A numerical model was developed that allows the prediction of the surface temperature distribution that is

  18. BWR Services maintenance training program

    International Nuclear Information System (INIS)

    Cox, J.H.; Chittenden, W.F.

    1979-01-01

    BWR Services has implemented a five-phase program to increase plant availability and capacity factor in operating BWR's. One phase of this program is establishing a maintenance training program on NSSS equipment; the scope encompasses maintenance on both mechanical equipment and electrical control and instrumentation equipment. The program utilizes actual product line equipment for practical Hands-on training. A total of 23 formal courses will be in place by the end of 1979. The General Electric Company is making a multimillion dollar investment in facilities to support this training. These facilities are described

  19. Thermal analyses for the spend fuel pool of Taiwan BWR plants during the loss of cooling accident

    Energy Technology Data Exchange (ETDEWEB)

    Chen, B-Y.; Yeh, C-L.; Wei, W-C.; Chen, Y-S., E-mail: onepicemine@iner.gov.tw, E-mail: clinyeh@iner.gov.tw, E-mail: hn150456@iner.gov.tw, E-mail: yschen@iner.gov.tw [Inst. of Nuclear Energy Research, Longtan Township, Taoyuan County, Taiwan (China)

    2014-07-01

    After the Fukushima nuclear accident, the safety of the spent fuel pool has become an important concern. In this study, thermal analysis of the spent fuel pool under a loss of cooling accident is performed. The BWR spent fuel pools in Taiwan are investigated, including the Chinshan, Kuosheng, and Lungmen plants. The transient pool temperature and level behaviors are calculated based on lumped energy balance. After the pool level drops below the top of the fuel, the peak cladding temperature is predicted by the Computational Fluid Dynamics (CFD) analysis. The influence to the cladding temperature of the uniform and checkboard fuel loading patterns is also investigated. (author)

  20. Study on the influence of water chemistry on fuel cladding behaviour of LWR in Japan

    International Nuclear Information System (INIS)

    Mishima, Y.

    1983-01-01

    This article presents the results of the study on the influence of water chemistry on fuel cladding behaviour, which has been performed for more than ten years on BWRs and PWRs in Japan. The post irradiation examination (P.I.E.) program of commercial reactor fuel assembly which was explained at Tokyo meeting in 1981 includes an investigation of the characteristics and build-up conditions of crud deposited on mainly BWR fuel cladding. This article also provides a summary of the results of the investigation and shows how the results are utilized for establishing effective water chemistry measures

  1. Stone cladding engineering

    National Research Council Canada - National Science Library

    Camposinhos, Rui de Sousa

    2014-01-01

    .... Straightforward formulae are provided for computing action on cladding, with special emphasis on the effect of seismic forces, including an extensive general methodology applied to non-structural elements...

  2. Cladding creepdown under compression

    International Nuclear Information System (INIS)

    Hobson, D.O.

    1977-01-01

    Light-water power reactors use Zircaloy tubing as cladding to contain the UO 2 fuel pellets. In-service operating conditions impose an external hydrostatic force on the cladding, causing it to creep down into eventual contact with the fuel. Knowledge of the rate of such creepdown is of great importance to modelers of fuel element performance. An experimental system was devised for studying creepdown that meets several severe requirements by providing (1) correct stress state, (2) multiple positions for measuring radial displacement of the cladding surface, (3) high-precision data, and (4) an experimental configuration compact enough to fit in-reactor. A microcomputer-controlled, eddy-current monitoring system was developed for this study and has proven highly successful in measuring cladding deformation with time at temperatures of 371 0 C (700 0 F) and higher, and at pressures as high as 21 MPa

  3. MELCOR 1.8.2 assessment: The DF-4 BWR Damaged Fuel experiment

    International Nuclear Information System (INIS)

    Tautges, T.J.

    1993-10-01

    MELCOR is a fully integrated, engineering-level computer code being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRs. As a part of an ongoing assessment, program, MELCOR has been used to model the ACRR in-pile DF-4 Damaged Fuel experiment. DF-4 provided data for early phase melt progression in BWR fuel assemblies, particularly for phenomena associated with eutectic interactions in the BWR control blade and zircaloy oxidation in the canister and cladding. MELCOR provided good agreement with experimental data in the key areas of eutectic material behavior and canister and cladding oxidation. Several shortcomings associated with the MELCOR modeling of BWR geometries were found and corrected. Twenty-five sensitivity studies were performed on COR, HS and CVH parameters. These studies showed that the new MELCOR eutectics model played an important role in predicting control blade behavior. These studies revealed slight time step dependence and no machine dependencies. Comparisons made with the results from four best-estimate codes showed that MELCOR did as well as these codes in matching DF-4 experimental data

  4. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    International Nuclear Information System (INIS)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F.

    2000-01-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  5. Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

    Energy Technology Data Exchange (ETDEWEB)

    Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F

    2000-07-01

    The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

  6. IFPE/IFA-508 and 515, PCMI Behaviour of Thin Cladding Rods, JAERI and HRP

    International Nuclear Information System (INIS)

    2007-01-01

    Description: To measure the integrated response of UO 2 and its cladding to conditions associated with PCI, the Japan Atomic Energy Research Institute carried out a series of experiments in the Halden BWR. The experiment involved two major objectives. The first was to study the influence of rod design parameters on PCI. Diametral gap, wall cladding thickness, SiO 2 additive, and pellet grain size were used as design parameters. The second objective was to study the influence of pre-irradiation (i.e. burnup) on PCI. The maximum burnup attained in the experiment was 23 MWd/kgU. These research results can be applied to current BWR-type fuel rods. The tests were performed between April 1977 and March 1981

  7. Development of alternative materials for BWR fuel springs

    International Nuclear Information System (INIS)

    Uruma, Y.; Osato, T.; Yamazaki, K.

    2002-01-01

    Major sources of radioactivity introduced into reactor water of BWR were estimated fuel crud and in-core materials (especially, fuel springs). Fuel springs are used for fixation of fuel cladding tubes with spacer grid. Those are small parts (total length is only within 25 mm) and so many numbers are loaded simultaneously and then total surfaces area are calculated up to about 200 m 2 . Fuel springs are located under high radiation field and high oxidative environment. Conventional fuel spring is made of alloy-X750 which is one of nickel-based alloy and is reported to show relatively higher corrosion release rate. 58 Co and 60 Co will be released directly into reactor water from intensely radio-activated fuel springs surface and increase radioactivity concentrations in primary coolant. Corrosion release control from fuel springs is an important technical item and a development of alternative material instead of alloy-X750 for fuel spring is a key subject to achieve ultra low man-rem exposure BWR plant. In present work, alloy-X718 which started usage for PWR fuel springs and stainless steel type 316L which has many mechanical property data are picked up for alternative materials and compared their corrosion behaviors with conventional material. Corrosion experiment was conducted under vapor-water two phases flow which is simulated fuel cladding surface boiling condition. After exposure, corrosion film formed under corrosion test was analyzed in detail and corrosion film amount and corrosion release amount are estimated among three materials. (authors)

  8. BWR and PWR chemistry operating experience and perspectives

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

    2014-01-01

    It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

  9. The possibility and the effects of a steam explosion in the BWR lower head on recriticality of a BWR core

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.

    2002-12-01

    The report describes an analysis considering a BWR postulated severe accident scenario during which the late vessel automatic depressurization brings the water below the level of the bottom core plate. The subsequent lack of ECCS leads to core heat up during which the control rods melt and the melt deposits on the core plate. At that point of time in the scenario, the core fuel bundles are still intact and the Zircaloy clad oxidation is about to start. The objective of the study is to provide the conditions of reflood into the hot core due to the level swell or a slug delivered from the lower head as the control rod melt drops into the water. These conditions are employed in the neutronic analysis with the RECRIT code to determine if the core recriticality may be achieved. (au)

  10. BWR alloy 182 stress Corrosion Cracking Experience

    International Nuclear Information System (INIS)

    Horn, R.M.; Hickling, J.

    2002-01-01

    Modern Boiling Water Reactors (BWR) have successfully operated for more than three decades. Over that time frame, different materials issues have continued to arise, leading to comprehensive efforts to understand the root cause while concurrently developing different mitigation strategies to address near-term, continued operation, as well as provide long-term paths to extended plant life. These activities have led to methods to inspect components to quantify the extent of degradation, appropriate methods of analysis to quantify structural margin, repair designs (or strategies to replace the component function) and improved materials for current and future application. The primary materials issue has been the occurrence of stress corrosion cracking (SCC). While this phenomenon has been primarily associated with austenitic stainless steel, it has also been found in nickel-base weldments used to join piping and reactor internal components to the reactor pressure vessel consistent with fabrication practices throughout the nuclear industry. The objective of this paper is to focus on the history and learning gained regarding Alloy 182 weld metal. The paper will discuss the chronology of weld metal cracking in piping components as well as in reactor internal components. The BWR industry has pro-actively developed inspection processes and procedures that have been successfully used to interrogate different locations for the existence of cracking. The recognition of the potential for cracking has also led to extensive studies to understand cracking behavior. Among other things, work has been performed to characterize crack growth rates in both oxygenated and hydrogenated environments. The latter may also be relevant to PWR systems. These data, along with the understanding of stress corrosion cracking processes, have led to extensive implementation of appropriate mitigation measures. (authors)

  11. Fundamental metallurgical aspects of axial splitting in zircaloy cladding

    International Nuclear Information System (INIS)

    Chung, H. M.

    2000-01-01

    Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10 21 n cm -2 to 5.9 x 10 21 n cm -2 (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed

  12. Laser cladding of turbine blades

    International Nuclear Information System (INIS)

    Shepeleva, L.; Medres, B.; Kaplan, W.D.; Bamberger, M.

    2000-01-01

    A comparative study of two different techniques for the application of wear-resistant coatings for contact surfaces of shroud shelves of gas turbine engine blades (GTE) has been conducted. Wear-resistant coatings were applied on In713 by laser cladding with direct injection of the cladding powder into the melt pool. Laser cladding was conducted with a TRUMPF-2500, CW-CO 2 laser. The laser cladding was compared with commercially available plasma cladding with wire. Both plasma and laser cladded zones were characterized by optical and scanning electron microscopy. It was found that the laser cladded zone has a higher microhardness value (650-820 HV) compared with that of the plasma treated material (420-440 HV). This is a result of the significant reduction in grain size in the case of laser cladding. Unlike the plasma cladded zones, the laser treated material is free of micropores and microcracks. (orig.)

  13. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Tsuchiya, Toshio; Masuda, Hisao; Isono, Tomoyuki; Noji, Kunio; Togo, Toshiki

    1989-01-01

    BWR Operator Training Center Corporation (BTC) was established in April 1971 for the purpose of training the operators from all BWR utilities in Japan. Since April 1974, more than 2600 operators and 1000 shift teams have been trained with the full-scope simulators in BTC up to the end of March 1988. To get the satisfactory results of the training, BTC has been making every effort to improve the facilities, the training materials, the instruction methods and the curricula. In this paper, such a series of recent improvements in the instruction methods and the curricula are presented that are effective to expand the knowledge and to improve the skills of middle or senior class operators. (author)

  14. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  15. BWR control blade replacement strategies

    Energy Technology Data Exchange (ETDEWEB)

    Kennard, M W [Stoller Nuclear Fuel, NAC International, Pleasantville, NY (United States); Harbottle, J E [Stoller Nuclear Fuel, NAC International, Thornbury, Bristol (United Kingdom)

    2000-02-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B{sub 4}C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  16. BWR control blade replacement strategies

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    The reactivity control elements in a BWR, the control blades, perform three significant functions: provide shutdown margin during normal and accident operating conditions; provide overall core reactivity control; and provide axial power shaping control. As such, the blades are exposed to the core's neutron flux, resulting in irradiation of blade structural and absorber materials. Since the absorber depletes with time (if B 4 C is used, it also swells) and the structural components undergo various degradation mechanisms (e.g., embrittlement, corrosion), the blades have limits on their operational lifetimes. Consequently, BWR utilities have implemented strategies that aim to maximize blade lifetimes while balancing operational costs, such as extending a refuelling outage to shuffle high exposure blades. This paper examines the blade replacement strategies used by BWR utilities operating in US, Europe and Asia by assembling information related to: the utility's specific blade replacement strategy; the impact the newer blade designs and changes in core operating mode were having on those strategies; the mechanical and nuclear limits that determined those strategies; the methods employed to ensure that lifetime limits were not exceeded during operation; and blade designs used (current and replacement blades). (author)

  17. Development of next BWR plant

    International Nuclear Information System (INIS)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke

    1995-01-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.)

  18. Development of next BWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Moriya, Kumiaki; Tanikawa, Naoshi; Kinoshita, Shoichiro; Utena, Shunsuke [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

    1995-04-01

    It is expected that BWR power generation will be main nuclear power generation for long period hereafter, and in the ABWRs being constructed at present, the safety, reliability, operation performance, economical efficiency and so on are further heightend as compared with conventional BWRs. On the other hand, in order to cope with future social change, the move to develop the next reactor type following ABWRs was begun already by the cooperation of electirc power companies and plant manufacturers. Hitachi Ltd. has advanced eagerly the development of new light water reactors. Also the objective of BWR power generation hereafter is to heighten the safety, reliability, operation performance and economical efficiency, and the development has been advanced, aiming at bearing the main roles of nuclear power generation. At present, ABWRs are under construction as No. 6 and 7 plants in Kashiwazaki Kariwa Nuclear Power Station, Tokyo Electric Power Co., Inc. In order to let ABWRs take root, the further improvement of economy by the standardization, the rationalization by revising the specification and the improvement of machinery and equipment is necessary. As the needs of the development of next generation BWRs, the increase of power output, the heightening of safety and economical efficiency are discussed. The concept of the next generation BWR plant aiming at the start of operation around 2010 is shown. (K.I.).

  19. Diffusion in cladding materials

    International Nuclear Information System (INIS)

    Anand, M.S.; Pande, B.M.; Agarwala, R.P.

    1992-01-01

    Aluminium has been used as a cladding material in most research reactors because its low neutron absorption cross section and ease of fabrication. However, it is not suitable for cladding in power reactors and as such zircaloy-2 is normally used as a clad because it can withstand high temperature. It has low neutron absorption cross section, good oxidation, corrosion, creep properties and possesses good mechanical strength. With the passage of time, further development in this branch of science took place and designers started looking for better neutron economy and less hydrogen pickup in PHW reactors. The motion of fission products in the cladding material could pose a problem after long operation. In order to understand their behaviour under reactor environment, it is essential to study first the diffusion under normal conditions. These studies will throw light on the interaction of defects with impurities which would in turn help in understanding the mechanism of diffusion. In this article, it is intended to discuss the diffusion behaviour of impurities in cladding materials.(i.e. aluminium, zircaloy-2, zirconium-niobium alloy etc.). (author). 94 refs., 4 figs., 3 tabs

  20. FEMAXI-7 analysis on behavior of medium and high burnup BWR fuels during base-irradiation and power ramp

    Energy Technology Data Exchange (ETDEWEB)

    Ogiyanagi, Jin, E-mail: ohgiyanagi.jin@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hanawa, Satoshi; Suzuki, Motoe; Nagase, Fumihisa [Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two power ramp experiments of BWR fuels were analyzed by FEMAXI-7 code. Black-Right-Pointing-Pointer Calculated FGR and cladding deformation showed reasonable agreement with PIE data. Black-Right-Pointing-Pointer High temperature FGR could be predicted by the enhanced Turnbull FG diffusion constant. Black-Right-Pointing-Pointer Local PCMI model in the code could reasonably predict cladding ridging deformation. - Abstract: Irradiation behavior of medium and high burnup BWR fuels during base-irradiation and subsequent power ramp test is analyzed by a fuel performance code FEMAXI-7. The code has a 1.5-D cylindrical geometry (4 axial segments) to have a coupled solution of thermal analysis and FEM mechanical analysis. Two kinds of target fuels are selected; one was subjected to a power ramp test in the DR3 reactor at RISO after the base-irradiation in a commercial BWR, and the other was subjected to the power ramp test in the DR3 reactor after the base-irradiation in the Halden boiling water reactor. The calculated values such as fission gas release after the base-irradiation and a cladding diameter profile before and after the ramp test show a reasonable agreement with measured data. In addition, the calculated ridging deformation of the cladding before and after the ramp test, which is obtained by using a local pellet-cladding mechanical interaction (PCMI) analysis geometry in FEMAXI-7, is compared with the measured data, and it is found that the FEMAXI-7 code is applicable to the local PCMI analysis of medium and high burnup rods under normal operation and power ramp conditions.

  1. BWR/5 Pressure-Suppression Pool Response during an SBO

    Directory of Open Access Journals (Sweden)

    Javier Ortiz-Villafuerte

    2013-01-01

    Full Text Available RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liquid water inventory in the PSP is vaporized. During the SBO, without RCIC, this inventory is reduced about 5% more within six hours. Further, a significant pressure rise occurs in containment at about the time when a sharp increase of heat generation occurs in RPV due to cladding oxidation. Failure temperature of fuel clad is also reached at this point. As the accident progresses, conditions for containment venting can be reached in about nine hours, although there still exists considerable margin before reaching containment design pressure. Detailed information of accident progress in reactor vessel and containment is presented and discussed.

  2. Electra-Clad

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-05-04

    The study relates to the use of building-integrated photovoltaics. The Electra-Clad project sought to use steel-based cladding as a substrate for direct fabrication of a fully integrated solar panel of a design similar to the ICP standard glass-based panel. The five interrelated phases of the project are described. The study successfully demonstrated that the principles of the panel design are achievable and sound. But, despite intensive trials, a commercially realistic solar performance has not been achieved: the main failing was the poor solar conversion efficiency as the active area of the panel was increased in size. The problem lies with the coating used on the steel cladding substrates and it was concluded that a new type of coating will be required. ICP Solar Technologies UK carried out the work under contract to the DTI.

  3. Nuclear fuel cladding material

    International Nuclear Information System (INIS)

    Nakahigashi, Shigeo.

    1982-01-01

    Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. (Sekiya, K.)

  4. Observations of crud deposits, corrosion and erosion of BWR and PWR fuel

    International Nuclear Information System (INIS)

    Bairiot, H.

    1983-01-01

    The BWR experience is limited to one reactor but the PWR experience covers a wide range of successive generations of power plants (7 in total). The systems are described and their water chemistry briefly commented. Some R and D performed on the effects of the operating regimes (steady state and transients) are summarized. Observations made by pool-side inspections and postirradiation examinations of fuel are outlined concerning water chemistry effects (crud deposits and corrosion) and ''mechanical'' coolant-cladding interaction (chip deposits and baffle jetting). (author)

  5. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  6. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  7. Sphere-pac versus pellet UO2 fuel in de Dodewaard BWR

    International Nuclear Information System (INIS)

    Linde, A. van der.

    1989-04-01

    Comparative testing of UO 2 sphere-pac and pellet fuel rods under LWR conditions has been jointly performed by the Netherlands Utilities Research Centre (KEMA) in Arnhem, the Netherlands Energy Research Foundation (ECN) at Petten and the Netherlands Joint Nuclear Power Utility (GKN) at Dodewaard. This final report summarizes the highlights of this 1968-1988 program with strong emphasis on the fuel rods irradiated in the Dodewaard BWR. The conclusion reached is that under normal LWR conditions sphere-pac UO 2 in LWR fuel rods offers better resistance against stress corrosion cracking of the cladding, but that under fast, single step, power ramping conditions pellet UO 2 in LWR fuel rods has a better resistance against hoop stress failure of the cladding. 128 figs., 36 refs., 19 tabs

  8. Studies on the fission products behavior during dissolution process of BWR spent fuel

    International Nuclear Information System (INIS)

    Sato, K.; Nakai, E.; Kobayashi, Y.

    1987-01-01

    In order to obtain basic data on fission products behavior in connection with the head end process of fuel reprocessing, especially to obtain better understanding on undissolved residues, small scale dissolution studies were performed by using BWR spent fuel rods which were irradiated as monitoring fuel rods under the monitoring program for LWR fuel assembly performance entitled PROVING TEST ON RELIABILITY OF FUEL ASSEMBLY . The Zircaloy-2 claddings and the fuel pellets were subjected individually to the following studies on 1) release of fission products during dissolution process, 2) characterization of undissolved residues, and 3) analysis of the claddings. This paper presents comprehensive descriptions of the fission products behavior during dissolution process, based on detailed and through PIE conducted by JNFS under the sponsorship of MITI (Ministry of International Trade and Industry)

  9. Arisings of cladding wastes from nuclear fuel in the European Community

    International Nuclear Information System (INIS)

    Cottone, G.

    1978-01-01

    An inquiry has been made in the member states on composition, activation and amounts of cladding wastes arising in the European Community until 1990 from the following reactor types: BWR, PWR, SGHWR, AGR and FBR. The elaborated results of this inquiry are given in this report. On the basis of forecasted reprocessing capacities the cumulative amount of cladding waste in the Community was estimated to reach in 1985 and 1990, respectively, about 2,100 and 7,300 metric tons. This waste will mainly consist of zircaloy and of smaller amounts of stainless steel and nickel alloy. Assuming that 0.5% of the spent fuel remains with the cladding, the contamination has been estimated for cooling times varying from 1 to 1000 years. In the first centuries activation is prevailing, but contamination determines the long-term radioactivity; consequently better decontamination, removing the alpha-bearing compounds, would be beneficial in reducing the long term hazard

  10. Study on transport safety of fresh MOX fuel. Performance of the cladding tube of fresh MOX fuel against external water pressure

    International Nuclear Information System (INIS)

    Ito, Chihiro

    1999-01-01

    It is important to know the ability of the cladding tube for fresh MOX fuel against external water pressure when they were hypothetically sunk into the sea for unknown reasons. In order to evaluate the ability of cladding tubes for MOX fresh fuel against external water pressure, external water pressure tests were carried out. Resistible limit of cladding tubes against external water pressure is defined when cladding tubes are deformed largely due to buckling etc. The test results show cladding tube of BWR type can resist an external water pressure of 69 MPa (a depth of water of 7,000 m) and that of PWR type fuel can resist an external water pressure of 54 MPa (a depth of water of 5,500 m). Moreover, leak tightness is maintained at an external water pressure of 73 MPa (a depth of water of 7,400 m) for BWR type cladding tubes and at an external water pressure of 98 MPa (a depth of water of 10,000 m) for PWR type cladding tubes. (author)

  11. 3D modeling of missing pellet surface defects in BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W., E-mail: Benjamin.Spencer@inl.gov; Williamson, R.L.; Stafford, D.S.; Novascone, S.R.; Hales, J.D.; Pastore, G.

    2016-10-15

    Highlights: • A global/local analysis procedure for missing pellet surface defects is proposed. • This is applied to defective BWR fuel under blade withdrawal and high power ramp conditions. • Sensitivity of the cladding response to key model parameters is studied. - Abstract: One of the important roles of cladding in light water reactor fuel rods is to prevent the release of fission products. To that end, it is essential that the cladding maintain its integrity under a variety of thermal and mechanical loading conditions. Local geometric irregularities in fuel pellets caused by manufacturing defects known as missing pellet surfaces (MPS) can in some circumstances lead to elevated cladding stresses that are sufficiently high to cause cladding failure. Accurate modeling of these defects can help prevent these types of failures. The BISON nuclear fuel performance code developed at Idaho National Laboratory can be used to simulate the global thermo-mechanical fuel rod behavior, as well as the local response of regions of interest, in either 2D or 3D. In either case, a full set of models to represent the thermal and mechanical properties of the fuel, cladding and plenum gas is employed. A procedure for coupling 2D full-length fuel rod models to detailed 3D models of the region of the rod containing a MPS defect is detailed here. The global and local model each contain appropriate physics and behavior models for nuclear fuel. This procedure is demonstrated on a simulation of a boiling water reactor (BWR) fuel rod containing a pellet with an MPS defect, subjected to a variety of transient events, including a control blade withdrawal and a ramp to high power. The importance of modeling the local defect using a 3D model is highlighted by comparing 3D and 2D representations of the defective pellet region. Parametric studies demonstrate the effects of the choice of gaseous swelling model and of the depth and geometry of the MPS defect on the response of the cladding

  12. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Ott, L.J.

    1997-01-01

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B 4 C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  13. Application of laser cladding method to small-diameter stainless steel pipes in actual nuclear plant

    International Nuclear Information System (INIS)

    Atago, Y.; Yamadera, M.; Tsuji, H.; Shiraiwa, T.; Kanno, M.

    1995-01-01

    Recently, to prevent stress corrosion cracking (SCC) the material of stainless steel (Type 304), a laser cladding method which produces a highly corrosion-resisting coating (cladding) to be formed on the surface of the material was developed. This is applicable to a long distance and narrow space, because of the good accessibility of the YAG (Yttrium-Aluminum Garnet) laser beam that can be transmitted through an optical fiber. In this method, a paste mixed metallic powder and heating resistive organic solvent is firstly placed on the inner surface of a small pipe and then a YAG laser beam transmitted through an optical fiber is irradiated to the paste, which will be melted and formed a clad subsequently, which is excellent in corrosion resistance. Finally, it can be achieved further resistance against the SCC due to the clad layer formed thus on the surface of the material. Recently, this Laser Cladding method was practically and successfully applied to the actual BWR Nuclear Power Plant in Japan. This report introduces the laser cladding technique, the equipments developed for practical application in the field

  14. BWR water chemistry impurity studies

    International Nuclear Information System (INIS)

    Ljungberg, L.G.; Korhonen, S.; Renstroem, K.; Hofling, C.G.; Rebensdorff, B.

    1990-03-01

    Laboratory studies were made on the effect of water impurities on environmental cracking in simulated BWR water of stainless steel, low alloy steel and nickel-base alloys. Constant elongation rate tensile (CERT) tests were run in simulated normal water chemistry (NWC), hydrogen water chemistry (HWC), or start-up environment. Sulfate, chloride and copper with chloride added to the water at levels of a fraction of a ppM were found to be extremely deleterious to all kinds of materials except Type 316 NG. Other detrimental impurities were fluoride, silica and some organic acids, although acetic acid was beneficial. Nitrate and carbon dioxide were fairly inoccuous. Corrosion fatigue and constant load tests on compact tension specimens were run in simulated normal BWR water chemistry (NWC) or hydrogen water chemistry (HWC), without impurities or with added sulfate or carbon dioxide. For sensitized Type 304 SS in NWC, 0.1 ppM sulfate increased crack propagation rates in constant load tests by up to a factor of 100, and in fatigue tests up to a factor of 10. Also, cracking in Type 316 nuclear grade SS and Alloy 600 was enhanced, but to a smaller degree. Carbon dioxide was less detrimental than sulfate. 3 figs., 4 tabs

  15. BWR core melt progression phenomena: Experimental analyses

    International Nuclear Information System (INIS)

    Ott, L.J.

    1992-01-01

    In the BWR Core Melt in Progression Phenomena Program, experimental results concerning severe fuel damage and core melt progression in BWR core geometry are used to evaluate existing models of the governing phenomena. These include control blade eutectic liquefaction and the subsequent relocation and attack on the channel box structure; oxidation heating and hydrogen generation; Zircaloy melting and relocation; and the continuing oxidation of zirconium with metallic blockage formation. Integral data have been obtained from the BWR DF-4 experiment in the ACRR and from BWR tests in the German CORA exreactor fuel-damage test facility. Additional integral data will be obtained from new CORA BWR test, the full-length FLHT-6 BWR test in the NRU test reactor, and the new program of exreactor experiments at Sandia National Laboratories (SNL) on metallic melt relocation and blockage formation. an essential part of this activity is interpretation and use of the results of the BWR tests. The Oak Ridge National Laboratory (ORNL) has developed experiment-specific models for analysis of the BWR experiments; to date, these models have permitted far more precise analyses of the conditions in these experiments than has previously been available. These analyses have provided a basis for more accurate interpretation of the phenomena that the experiments are intended to investigate. The results of posttest analyses of BWR experiments are discussed and significant findings from these analyses are explained. The ORNL control blade/canister models with materials interaction, relocation and blockage models are currently being implemented in SCDAP/RELAP5 as an optional structural component

  16. SCORPIO-BWR: status and future plans

    International Nuclear Information System (INIS)

    Porsmyr, Jan; Bodal, Terje; Beere, William H.

    2004-01-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR system on a

  17. SCORPIO-BWR: status and future plans

    Energy Technology Data Exchange (ETDEWEB)

    Porsmyr, Jan; Bodal, Terje; Beere, William H. (and others)

    2004-07-01

    Full text: During the years from 2000 to 2003 a joint project has been performed by IFE, Halden and TEPCO Systems Corporation, Japan, to develop a core monitoring system for BWRs based on the their existing core monitoring system TiARA and the SCORPIO framework. It has been emphasised to develop a reliable, flexible, adaptable and user-friendly system, which is easy to maintain. Therefore, a rather general framework (SCORPIO Framework) has been used which facilitates easy software modifications as well as adding/ replacing physics modules. The software modules is integrated in the SCORPIO framework using the Software Bus as the communication tool and with the Picasso UIMS tool for MMI. The SCORPIO-BWR version is developed on a Windows-PC platform. The SCORPIO-BWR version provides all functions, which are necessary for all analyses and operations performed on a BWR plant and comprises functions for on-line core monitoring, predictive analysis and core management with interfaces to plant instrumentation and physics codes. Functions for system initialisation and maintenance are also included. A SCORPIO-BWR version adapted for ABWR was installed in TEPSYS facilities in Tokyo in January 2003, where the final acceptance tests were carried out and accepted. The ABWR version of the system is now in the verification and validation phase. In the period from April 2003 until March 2004 a project for realizing an offline-version of SCORPIO-BWR system, which supports the offline tasks of BWR in-core fuel management for ABWR and BWR-5 type of reactors, was developed. The offline-version of the SCORPIO-BWR system for ABWR and BWR-5 type of reactors was installed at TEPSYS in March 2003, where the final acceptance tests were carried out and accepted. Plans for the next version of this system is to study the possibility of adapting SCORPIO-BWR to work with 'mobile technology'. This means that it should be possible to access and display information from the SCORPIO-BWR

  18. Advances in BWR water chemistry

    International Nuclear Information System (INIS)

    Garcia, Susan E.; Giannelli, Joseph F.; Jarvis, Mary L.

    2012-09-01

    This paper reviews recent advances in Boiling Water Reactor (BWR) water chemistry control with examples of plant experiences at U.S. designed BWRs. Water chemistry advances provide some of the most effective methods for mitigating materials degradation, reducing fuel performance concerns and lowering radiation fields. Mitigation of stress corrosion cracking (SCC) of materials remains a high priority and improved techniques that have been demonstrated in BWRs will be reviewed, specifically hydrogen injection combined with noble metal chemical addition (NMCA) and the newer on-line noble metal application process (OLNC). Hydrogen injection performance, an important part of SCC mitigation, will also be reviewed for the BWR fleet, highlighting system improvements that have enabled earlier injection of hydrogen including the potential for hydrogen injection during plant startup. Water chemistry has been significantly improved by the application of pre-filtration and optimized use of ion exchange resins in the CP (condensate polishing) and reactor water cleanup (RWCU) systems. EPRI has monitored and supported water treatment improvements to meet water chemistry goals as outlined in the EPRI BWR Water Chemistry Guidelines, particularly those for SCC mitigation of reactor internals and piping, minimization of fuel risk due to corrosion and crud deposits and chemistry control for radiation field reduction. In recent years, a significant reduction has occurred in feedwater corrosion product input, particularly iron. A large percentage of plants are now reporting <0.1 ppb feedwater iron. The impacts to plant operation and chemistry of lower feedwater iron will be explored. Depleted zinc addition is widely practiced across the fleet and the enhanced focus on radiation reduction continues to emphasize the importance of controlling radiation source term. In addition, shutdown chemistry control is necessary to avoid excessive release of activated corrosion products from fuel

  19. Investigation of 3H and 14C inventory and distribution in spent BWR fuel rods

    International Nuclear Information System (INIS)

    Bleier, A.; Beuerle, M.; Neeb, K.H.

    1984-10-01

    In order to obtain reliable data for fuel reprocessing and waste disposal, the T and C-14 inventory, distribution and behaviour was investigated on a typical LWR fuel rod discharged from a BWR plant. The results showed that 50 ± 5% of the T generated in the fuel is present in the cladding after reactor operation. The remainder of the T stays with the fuel. Related to the reactor power the total T inventory corresponds to a T production rate of 19 000 Ci/GW e . a. The C-14 built up in the fuel represents approximately 60% of the C-14 inventory of the BWR fuel rod. The remaining part of C-14 (about 40%) experimentally determined by this analysis for the first time is generated in the cladding. From the total C-14 inventory a C-14 production rate of 17,5 Ci/GW e . a can be calculated. The fill gas contains only negligible fractions of both nuclides. The results obtained in this program are generally in good agreement with the data of theoretical estimates and with results of earlier investigations on PWR fuel rods. (orig.) [de

  20. BWR Assembly Optimization for Minor Actinide Recycling

    Energy Technology Data Exchange (ETDEWEB)

    G. Ivan Maldonado; John M. Christenson; J.P. Renier; T.F. Marcille; J. Casal

    2010-03-22

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs).

  1. BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Matsumoto, Kosuke.

    1991-01-01

    In a BWR type nuclear power plant in which reactor water in a reactor pressure vessel can be drained to a waste processing system by way of reactor recycling pipeways and remaining heat removal system pipeways, a pressurized air supply device is disposed for supplying air for pressurizing reactor water to the inside of the reactor pressure vessel by way of an upper head. With such a constitution, since the pressurized air sent from the pressurized air supply device above the reactor pressure vessel for the reactor water discharging pressure upon draining, the water draining pressure is increased compared with a conventional case and, accordingly, the amount of drained water is not reduced even in the latter half of draining. Accordingly, the draining efficiency can be improved and only a relatively short period of time is required till the completion of the draining, which can improve safety and save labors. (T.M.)

  2. Development of a detailed BWR core thermal-hydraulic analysis method based on the Japanese post-BT standard using a best-estimate code

    International Nuclear Information System (INIS)

    Ono, H.; Mototani, A.; Kawamura, S.; Abe, N.; Takeuchi, Y.

    2004-01-01

    The post-BT standard is a new fuel integrity standard or the Atomic Energy Society of Japan that allows temporary boiling transition condition in the evaluation for BWR anticipated operational occurrences. For application of the post-BT standard to BWR anticipated operational occurrences evaluation, it is important to identify which fuel assemblies and which axial, radial positions of fuel rods have temporarily experienced the post-BT condition and to evaluates how high the fuel cladding temperature rise was and how long the dryout duration continued. Therefore, whole bundle simulation, in which each fuel assembly is simulated independently by one thermal-hydraulic component, is considered to be an effective analytical method. In the present study, a best-estimate thermal-hydraulic code, TRACG02, has been modified to extend it predictive capability by implementing the post-BT evaluation model such as the post-BT heat transfer correlation and rewetting correlation and enlarging the number of components used for BWR plant simulation. Based on new evaluation methods, BWR core thermal-hydraulic behavior has been analyzed for typical anticipated operational occurrence conditions. The location where boiling transition occurs and the severity of fuel assembly in the case of boiling transition conditions such as fuel cladding temperature, which are important factors in determining whether the reuse of the fuel assembly can be permitted, were well predicted by the proposed evaluation method. In summary, a new evaluation method for a detailed BWR core thermal-hydraulic analysis based on the post-BT standard of the Atomic Energy Society of Japan has been developed and applied to the evaluation of the post-BT standard during the actual BWR plant anticipated operational occurrences. (author)

  3. Cladding tube manufacturing technology

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report A lloy Development for High Burnup Cladding . Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs

  4. Cladding tube manufacturing technology

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report 'Alloy Development for High Burnup Cladding.' Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs.

  5. Protecting AREVA ATRIUM™ BWR fuel from debris fretting failure

    International Nuclear Information System (INIS)

    Cole, Steven E.; Garner, Norman L.; Lippert, Hans-Joachim; Graebert, Rüdiger; Mollard, Pierre; Hahn, Gregory C.

    2014-01-01

    Historically, debris fretting has been the leading cause of fuel rod failure in BWR fuel assemblies, costing the industry millions of dollars in lost generation and negatively impacting the working area of plant site personnel. In this paper the focus will be on recent BWR fuel product innovation designed to eliminate debris related failures. Experience feedback from more than three decades of operation history with non-line-of-sight FUELGUARD™ lower tie plate debris filters will be presented. The development and relative effectiveness of successive generations of filtration technology will be discussed. It will be shown that modern, state of the art debris filters are an effective defense against debris fretting failure. Protective measures extend beyond inlet nozzle debris filters. The comprehensive debris resistance features built into AREVA’s newest fuel design, the ATRIUM™ 11, reduce the overall risk of debris entrapment as well as providing a degree of protection from debris that may fall down on the fuel assembly from above, e.g., during refueling operations. The positive recent experience in a debris sensitive plant will be discussed showing that the combination of advanced fuel technology and a robust foreign material exclusion program at the reactor site can eliminate the debris fretting failure mechanism. (author)

  6. Simplified compact containment BWR plant

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Tsutagawa, M.; Hiraiwa, K.; Arai, K.; Hida, T.

    2004-01-01

    The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, is expected to prove attractive in the world energy markets due to its flexibility in regard to both energy demands and site conditions, its high potential for reducing investment risk and its safety features facilitating public acceptance. The flexibility is achieved by CCR's small power output of 300 MWe class and capability of long operating cycle (refueling intervals). CCR is expected to be attractive from view point of investment due to its simplification/innovation in design such as natural circulation core cooling with the bottom located short core, internal upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure containment concept. The natural circulation core eliminates recirculation pumps and the maintenance of such pumps. The internal upper entry CRDs reduce the height of the reactor vessel (RPV) and consequently reduce the height of the primary containment vessel (PCV). The safety features mainly consist of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), passive auto catalytic recombiner and in-vessel retention (IVR) capability. The large inventory increases the system response time in the case of design-base accidents, including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. The recombiner decreases hydrogen concentration in the PCV in the case of a severe accident. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. The feasibility of CCR safety system has been confirmed by LOCA

  7. CASTI handbook of cladding technology. 2. ed.

    International Nuclear Information System (INIS)

    Smith, L.; Celant, M.

    2000-01-01

    This updated (2000) CASTI handbook covers all aspects of clad products - the different means of manufacture, properties and applications in various industries. Topics include: an introduction to cladding technology, clad plate, clad pipes, bends, clad fittings, specification requirements of clad products, welding clad products, clad product application and case histories from around the world. Unique to this book is the documentation of case histories of major cladding projects from around the world and how the technology of that day has withstood the demands of time. Filled with over 100 photos and graphics illustrating the various cladding technology examples and products, this book truly documents the most recent technologies in the field of cladding technology used worldwide

  8. BISON Fuel Performance Analysis of FeCrAl cladding with updated properties

    Energy Technology Data Exchange (ETDEWEB)

    Sweet, Ryan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wirth, Brian [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-08-30

    operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.

  9. Improvement for BWR operator training

    International Nuclear Information System (INIS)

    Kurisu, Takanori; Takahashi, Yoshitaka; Harada, Mitsuhiro; Takahashi, Iwao.

    1988-01-01

    BWR Operator Training Center was founded in April, 1971, and in April, 1974, training was begun, since then, 13 years elapsed. During this period, the curriculum and training facilities were strengthened to meet the training needs, and the new training techniques from different viewpoint were developed, thus the improvement of training has been done. In this report, a number of the training techniques which have been developed and adopted recently, and are effective for the improvement of the knowledge and skill of operators are described. Recently Japanese nuclear power stations have been operated at stable high capacity factor, accordingly the chance of experiencing the occurrence of abnormality and the usual start and stop of plants decreased, and the training of operators using simulators becomes more important. The basic concept on training is explained. In the standard training course and the short period fundamental course, the development of the guide for reviewing lessons, the utilization of VTRs and the development of the techniques for diagnosing individual degree of learning were carried out. The problems, the points of improvement and the results of these are reported. (Kako, I.)

  10. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  11. Standard for assessment of fuel integrity under anticipated operational occurrences in BWR power plant:2002

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Suzuki, Riichiro; Komura, Seiichi; Kudo, Yoshiro; Yamanaka, Akihiro; Oomizu, Satoru; Kitamura, Hideya; Nagata, Yoshifumi

    2003-01-01

    To secure fuel integrity, a Light Water Reactor (LWR) core is designed so that no boiling transition (BT) should take place in fuel assemblies and excessive rise in fuel cladding temperature due to deteriorated that transfer should be avoided in Anticipated Operational Occurrences (AOO). In some AOO in a Boiling Water Reactor (BWR), however, the rise in reactor power could be limited by SCRAM or void reactivity effect. Recent studies have provided accumulated knowledge that even if BT takes place in fuel assemblies, the rise in fuel cladding temperature could be so small that it will not threat to fuel integrity, as long as the BT condition terminates within a short period of time. In addition, appropriate methods have been developed to evaluate the cladding temperature during dryout. This standard provides requirements in the assessment of fuel integrity under AOO in which limited-BT condition is temporarily reached and the propriety of reusing a fuel assembly that has experienced limited-BT condition. The standard has been approved by the Atomic Energy Society of Japan following deliberation by impartial members for two and half years. It is now expected that this standard will provide an effective measure for the rational expansion of fuel design and operational margin. (author)

  12. BWR Refill-Reflood Program. Final report

    International Nuclear Information System (INIS)

    Myers, L.L.

    1983-09-01

    The BWR Refill-Reflood Program is part of the continuing Loss of Coolant Accident (LOCA) research in the United States which is jointly sponsored by the Nuclear Regulatory Commission, the Electric Power Research Institute, and the General Electric Company. The current program expanded the focus of this research to include full scale experimental evaluations of multidimensional and multichannel effects during system refill. The program has also made major contributions to the BWR version of the Transient Reactor Analysis Code (TRAC) which has been developed cooperatively with the Idaho National Engineering Laboratory (INEL) for application to BWR transients. A summary description of the complete program is provided including the principal findings and main conclusions of the program. The results of the program have shown that multidimensional and parallel channel effects have the potential to significantly improve the system response over that observed in single channel tests

  13. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1983-01-01

    A survey is given of the main incentives for power reactor noise research, and the differences and similarities of noise in power and zero power systems are shown. After a short outline of historical developments the basic characteristics of the adjoint method in reactor noise theory are dealt with. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies, which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurements on a BWR in The Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  14. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  15. BWR radiation exposure--experience and projection

    International Nuclear Information System (INIS)

    Falk, C.F.; Wilkinson, C.D.; Hollander, W.R.

    1979-01-01

    The BWR/6 Mark III radiation exposures are projected to be about half of those of current average operating experience of 725 man-rem. These projections are said to be realistic and based on current achievements and not on promises of future development. The several BWRs operating with low primary system radiation levels are positive evidence that radiation sources can be reduced. Improvements have been made in reducing the maintenance times for the BWR/6, and further improvements can be made by further attention to cost-effective plant arrangement and layout during detail design to improve accessibility and maintainability of each system and component

  16. General Electric's training program for BWR chemists

    International Nuclear Information System (INIS)

    Osborn, R.N.; Lim, W.

    1981-01-01

    This paper describes the development and implementation of the General Electric boiling water reactor chemistry training program from 1959 to the present. The original intention of this program was to provide practical hands on type training in radiochemistry to BWR chemistry supervisors with fossil station experience. This emphasis on radiochemistry has not changed through the years, but the training has expanded to include the high purity water chemistry of the BWR and has been modified to include new commission requirements, engineering developments and advanced instrumentation. Student and instructor qualifications are discussed and a description of the spin off courses for chemistry technicians and refresher training is presented

  17. BWR plant analyzer development at BNL

    International Nuclear Information System (INIS)

    Cheng, H.S.; Wulff, W.; Mallen, A.N.; Lekach, S.V.; Stritar, A.; Cerbone, R.J.

    1985-01-01

    Advanced technology for high-speed interactive nuclear power plant simulations is of great value for timely resolution of safety issues, for plant monitoring, and for computer-aided emergency responses to an accident. Presented is the methodology employed at BNL to develop a BWR plant analyzer capable of simulating severe plant transients at much faster than real-time process speeds. Five modeling principles are established and a criterion is given for selecting numerical procedures and efficient computers to achieve the very high simulation speeds. Typical results are shown to demonstrate the modeling fidelity of the BWR plant analyzer

  18. The BWR Hybrid 4 control rod

    International Nuclear Information System (INIS)

    Gross, H.; Fuchs, H.P.; Lippert, H.J.; Dambietz, W.

    1988-01-01

    The service life of BWR control rods designed in the past has been unsatisfactory. The main reason was irradiation assisted stress corrosion cracking of B 4 C rods caused by external swelling of the B 4 C powder. By this reason KWU developed an improved BWR control rod (Hybrid 4 control rod) with extended service life and increased control rod worth. It also allows the procedure for replacing and rearranging fuel assemblies to be considerably simplified. A complete set of Hydbrid 4 control rods is expected to last throughout the service life of a plant (assumption: ca. 40 years) if an appropriate control rod reshuffling management program is used. (orig.)

  19. Improvement for BWR operator training, 3

    International Nuclear Information System (INIS)

    Noji, Kunio; Toeda, Susumu; Saito, Genhachi; Suzuki, Koichi

    1990-01-01

    BWR Operator Training Center Corporation (BTC) is conducting training for BWR plant operators using Full-scope Simulators. There are several courses for individual operators and one training course for shift crew (Family Training Course) in BTC. Family Training is carried out by all members of the operating shift-crew. BTC has made efforts to improve the Family Training in order to acquire more effective training results and contribute to up-grade team performance of all crews. This paper describes some items of our efforts towards Family Training improvement. (author)

  20. Panorama of the BWR reactors - Evolution of the concept

    Energy Technology Data Exchange (ETDEWEB)

    Novotny, C.; Uhrig, E. [AREVA NP GmbH, Safety Engineering Department - PEPS-G (Germany)

    2012-01-15

    Nowadays, a fleet of more than 50 boiling water reactors (BWR) are in operation in the world. This article gives a short overview on the developments of nuclear power plants of the BWR type, with a focus on the European builds. It describes the technical bases from the early designs in the fifties, sketches the innovations of the sixties and seventies in the types BWR 69 and 72 (Baulinie 69 and 72) and gives an outlook of a possible next generation BWR. A promising approach in recent BWR developments is the the combination of passive safety systems with established design basis

  1. Zirconium-barrier cladding attributes

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.; Rand, R.A.; Tucker, R.P.; Cheng, B.; Adamson, R.B.; Davies, J.H.; Armijo, J.S.; Wisner, S.B.

    1987-01-01

    This metallurgical study of Zr-barrier fuel cladding evaluates the importance of three salient attributes: (1) metallurgical bond between the zirconium liner and the Zircaloy substrate, (2) liner thickness (roughly 10% of the total cladding wall), and (3) softness (purity). The effect that each of these attributes has on the pellet-cladding interaction (PCI) resistance of the Zr-barrier fuel was studied by a combination of analytical model calculations and laboratory experiments using an expanding mandrel technique. Each of the attributes is shown to contribute to PCI resistance. The effect of the zirconium liner on fuel behavior during off-normal events in which steam comes in contact with the zirconium surface was studied experimentally. Simulations of loss-of-coolant accident (LOCA) showed that the behavior of Zr-barrier cladding is virtually indistinguishable from that of conventional Zircaloy cladding. If steam contacts the zirconium liner surface through a cladding perforation and the fuel rod is operated under normal power conditions, the zirconium liner is oxidized more rapidly than is Zircaloy, but the oxidation rate returns to the rate of Zircaloy oxidation when the oxide phase reaches the zirconium-Zircaloy metallurgical bond

  2. Moderator temperature coefficient in BWR core

    International Nuclear Information System (INIS)

    Naito, Yoshitaka

    1977-01-01

    Temperature dependences of infinite multiplication factor k sub(infinity) and neutron leakage from the core must be examined for estimation of moderator temperature coefficient. Temperature dependence on k sub(infinity) has been investigated by many researchers, however, the dependence on neutron leakage of a BWR with cruciformed control rods has hardly been done. Because there are difficulties and necessity on calculations of three space dimensional and multi-energy groups neutron distribution in a BWR core. In this study, moderator temperature coefficients of JPDR-II (BWR) core were obtained by calculation with DIFFUSION-ACE, which is newly developed three-dimensional multi-group computer code. The results were compared with experimental data measured from 20 to 275 0 C of the moderator temperature and the good agreement was obtained between calculation and measurement. In order to evaluate neutron leakage from the core, the other two calculations were carried out, adjusting criticality by uniform absorption rate and by material buckling. The former underestimated neutron leakage and the latter overestimated it. Discussion on the results shows that in order to estimate the temperature coefficient of BWR, neutron leakage must be evaluated precisely, therefore the calculation at actual pattern of control rods is necessary. (auth.)

  3. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  4. Delivering high performance BWR fuel reliably

    International Nuclear Information System (INIS)

    Schardt, J.F.

    1998-01-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  5. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  6. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    Candidate mitigative strategies for management of in-vessel events during the late phase (after core degradation has occurred) of postulated BWR severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for additional assessment. The first is a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertains to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose is to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies have been performed during 1991 under the auspices of the Detailed Assessment of BWR In-Vessel Strategies Program. This paper provides a discussion of the motivation for and purpose of these strategies and the potential for their success. 33 refs., 9 figs

  7. BWR stability analysis at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Rohatgi, U.S.

    1991-01-01

    Following the unexpected, but safely terminated, power and flow oscillations in the LaSalle-2 Boiling Water Reactor (BWR) on March 9, 1988, the Nuclear Regulatory Commission (NRC) Offices of Nuclear Reactor Regulation (NRR) and of Analysis and Evaluation of Operational Data (AEOD) requested that the Office of Nuclear Regulatory Research (RES) carry out BWR stability analyses, centered around fourteen specific questions. Ten of the fourteen questions address BWR stability issues in general and are dealt with in this paper. The other four questions address local, out-of-phase oscillations and matters of instrumentation; they fall outside the scope of the work reported here. It was the purpose of the work documented in this report to answer ten of the fourteen NRC-stipulated questions. Nine questions are answered by analyzing the LaSalle-2 instability and related BWR transients with the BNL Engineering Plant Analyzer (EPA) and by performing an uncertainty assessment of the EPA predictions. The tenth question is answered on the basis of first principles. The ten answers are summarized

  8. Characteristics of hydride precipitation and reorientation in spent-fuel cladding

    International Nuclear Information System (INIS)

    Chung, H.M.; Daum, R.S.; Hiller, J.M.; Billone, M.C.

    2002-01-01

    Transmission electron microscopy (TEM) was used to examine Zircaloy fuel cladding, either discharged from several PWRs and a BWR after irradiation to fluence levels of 3.3 to 8.6 X 10 21 n cm -2 (E > 1 MeV) or hydrogen-charged and heat-treated under stress to produce radial hydrides; the goal was to determine the microstructural and crystallographic characteristics of hydride precipitation. Morphologies, distributions, and habit planes of various types of hydrides were determined by stereo-TEM. In addition to the normal macroscopic hydrides commonly observed by optical microscopy, small 'microscopic' hydrides are present in spent-fuel cladding in number densities at least a few orders of magnitude greater than that of macroscopic hydrides. The microscopic hydrides, observed to be stable at least up to 333 deg C, precipitate in association with -type dislocations. While the habit plane of macroscopic tangential hydrides in the spent-fuel cladding is essentially the same as that of unirradiated unstressed Zircaloys, i.e., the [107] Zr plane, the habit plane of tangential hydrides that precipitate under high tangential stress is the [104] Zr plane. The habit plane of radial hydrides that precipitate under tangential stress is the [011] Zr pyramidal plane, a naturally preferred plane for a cladding that has 30 basal-pole texture. Effects of texture on the habit plane and the threshold stress for hydride reorientation are also discussed. (authors)

  9. Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide

    International Nuclear Information System (INIS)

    Gallardo V, J. M.; Morales S, J. B.

    2013-10-01

    The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO 2 ) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO 2 mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

  10. Clad Treatment in KARMA Code and Library

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong-yeup; Lee, Hae-chan; Woo, Hae-seuk [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of)

    2016-05-15

    Zirconium is the main components in clad materials. The subgroup parameters of zirconium were generated with effective cross section which obtained by using flux distribution in clad region. It decreases absorption reaction rate differences with reference MCNP results. Use of composite nuclide is acceptable to increase efficiency but should be limited to specific target composition. Therefore, the use of the composite nuclide of Zircaloy-2 should be limited when HANA clad material is used for clad. Either using explicit components or generating composite nuclide for HANA is suggested. This paper investigates the clad analysis model for KARMA whether current method is applicable to HANA clad material.

  11. Data report of BWR post-CHF tests. Transient core thermal-hydraulic test program. Contract research

    International Nuclear Information System (INIS)

    Iguchi, Tadashi; Itoh, Hideo; Kiuchi, Toshio; Watanabe, Hironori; Kimura, Mamoru; Anoda, Yoshinari

    2001-03-01

    JAERI has been performing transient core thermal-hydraulic test program. In the program, authors performed BWR/ABWR DBE simulation tests with a test facility, which can simulate BWR/ABWR transients. The test facility has a 4 x 4 bundle core simulator with 15-rod heaters and one non-heated rod. Through the tests, authors quantified the thermal safety margin for core cooling. In order to quantify the thermal safety margin, authors collected experimental data on post-CHF. The data are essential for the evaluation of clad temperature transient when core heat-up occurs during DBEs. In comparison with previous post-CHF tests, present experiments were performed in much wider experimental condition, covering high clad temperature, low to high pressure and low to high mass flux. Further, data at wider elevation (lower to higher elevation of core) were obtained in the present experiments, which make possible to discuss the effect of axial position on thermal-hydraulics, while previous works usually discuss the thermal-hydraulics at the position where the first heat-up occurs. This data report describes test procedure, test condition and major experimental data of post-CHF tests. (author)

  12. Infinite fuel element simulation of pin power distributions and control blade history in a BWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Li, J.; Nuenighoff, K.; Allelein, H.J. [Forschungszentrum Juelich GmbH (DE). Inst. fuer Energie- und Klimaforschung (IEK), Sicherheitsforschung und Reaktortechnik (IEK-6)

    2011-07-01

    Pellet-Cladding Interaction (PCI) is a well known effect in fuel pins. One possible reason for PCI-effects could be local power excursions in the fuel pins, which can led to a rupture of the fuel cladding tube. From a reactor safety point of view this has to be considered as a violence of the barrier principal in order to retain fission products in the fuel pins. This paper focuses on the pin power distributions in a 2D infinite lattice of a BWR fuel element. Lots of studies related PCI effect can be found in the literature. In this compact, coupled neutronic depletion calculations taking the control history effect into account are described. Depletion calculations of an infinite fuel element of a BWR were carried out with controlled, uncontrolled and temporarily controlled scenarios. Later ones are needed to describe the control blade history (CBH) effect. A Monte-Carlo approach is mandatory to simulate the neutron physics. The VESTA code was applied to couple the Monte-Carlo-Code MCNP(X) with the burnup code ORIGEN. Additionally, CASMO-4 is also employed to verify the method of simulation results from VESTA. The cross sections for Monte Carlo and burn-up calculations are derived from ENDF/B-VII.0. (orig.)

  13. Summary report of seismic PSA of BWR model plant

    International Nuclear Information System (INIS)

    1999-05-01

    This report presents a seismic PSA (Probabilistic Safety Assessment) methodology developed at the Japan Atomic Energy Research Institute (JAERI) for evaluating risks of nuclear power plants (NPPs) and the results from an application of the methodology to a BWR plant in Japan, which is termed Model Plant'. The seismic PSA procedures developed at JAERI are to evaluate core damage frequency (CDF) and have the following four steps: (1) evaluation of seismic hazard, (2) evaluation of realistic response, (3) evaluation of component capacities and failure probabilities, and (4) evaluation of conditional probability of system failure and CDF. Although these procedures are based on the methodologies established and used in the United States, they include several unique features: (1) seismic hazard analysis is performed with use of available knowledge and database on seismological conditions in Japan; (2) response evaluation is performed with a response factor method which is cost effective and associated uncertainties can be reduced with use of modern methods of design calculations; (3) capacity evaluation is performed with use of test results available in Japan in combination with design information and generic capacity data in the U.S.A.; (4) systems reliability analysis, performed with use of the computer code SECOM-2 developed at JAERI, includes identification of dominant accident sequences, importance analysis of components and systems as well as the CDF evaluation with consideration of the effect of correlation of failures by a newly developed method based on the Monte Carlo method. The effect of correlation has been recognized as an important issue in seismic PSAs. The procedures was used to perform a seismic PSA of a 1100 MWe BWR plant. Results are shown as well as the insights derived and future research needs identified in this seismic PSA. (J.P.N.)

  14. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    programs that GE has participated in and describes the different options and approaches that have been used by various utilities in their design basis programs. Some of these variations deal with the scope and depth of coverage of the information, while others are related to the process (how the work is done). Both of these topics can have a significant effect on the program cost. Some insight into these effects is provided. The final section of the paper presents a set of lessons learned and a recommendation for an optimum approach to a design basis information program. The lessons learned reflect the knowledge that GE has gained by participating in design basis programs with nineteen domestic and international BWR owner/operators. The optimum approach described in this paper is GE's attempt to define a set of information and a work process for a utility/GE NSSS Design Basis Information program that will maximize the cost effectiveness of the program for the utility. (author)

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

    International Nuclear Information System (INIS)

    Heuser, Brent

    2013-01-01

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

  16. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    E. Siegmann

    2004-01-01

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  17. Pin clad strains in Phenix

    International Nuclear Information System (INIS)

    Languille, A.

    1979-07-01

    The Phenix reactor has operated for 4 years in a satisfactory manner. The first 2 sub-assembly loadings contained pins clad in solution treated 316. The principal pin strains are: diametral strain (swelling and irradiation creep), ovality and spiral bending of the pin (interaction of wire and pin cluster and wrapper). A pin cluster irradiated to a dose of 80 dpa F reached a pin diameter strain of 5%. This strain is principally due to swelling (low fission gas pressure). The principal parameters governing the swelling are instantaneous dose, time and temperature for a given type of pin cladding. Other types of steel are or will be irradiated in Phenix. In particular, cold-worked titanium stabilised 316 steel should contribute towards a reduction in the pin clad strains and increase the target burn-up in this reactor. (author)

  18. Fuel-cladding chemical interaction

    International Nuclear Information System (INIS)

    Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J.

    2015-01-01

    The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels (oxide, carbide, nitride, metallic) to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. (authors)

  19. Influence of manufacturing process on the in-reactor creep anisotropy of stress-relieved Zircaloy-2 cladding

    International Nuclear Information System (INIS)

    Shann, S.H.; Van Swam, L.F.

    1995-01-01

    A procedure to determine the axial/radial and circumferential/radial contractile strain ratios (the R and P factors respectively in the Backofen-modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 cladding of BWR fuel rods, tie rods and water rods was developed and has been described previously (S.H. Shann and L.F. van Swam, Creep anisotropy of Zircaloy-2 cladding during irradiation, Trans. SMiRT-11, Vol. C, 1991). The present study employs the procedure to determine the anisotropy factors R and P for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding fabricated by various manufacturing processes. The analysis indicates that the cladding manufacturing process can have a pronounced effect on the anisotropy of irradiation-induced creep. Cladding types with identical yield and ultimate tensile strengths but fabricated by different manufacturing processes have different values of R and P during in-reactor creep. ((orig.))

  20. Development of a BWR core burn-up calculation code COREBN-BWR

    International Nuclear Information System (INIS)

    Morimoto, Yuichi; Okumura, Keisuke

    1992-05-01

    In order to evaluate core performances of BWR type reactors, the three dimensional core burnup calculation code COREBN-BWR and the fuel management code HIST-BWR have been developed. In analyses of BWR type reactors, thermal hydraulics calculations must be coupled with neutronics calculations to evaluate core performances, because steam void distribution changes according to the change of the power distribution. By installing new functions as follows to the three dimensional core burnup code COREBN2 developed in JAERI for PWR type reactor analyses, the code system becomes to be applicable to burnup analyses of BWR type reactors. (1) Macroscopic cross section calculation function taking into account of coolant void distribution. (2) Thermal hydraulics calculation function to evaluate core flow split, coolant void distribution and thermal margin. (3) Burnup calculation function under the Haling strategy. (4) Fuel management function to incorporate the thermal hydraulics information. This report consists of the general description, calculational models, input data requirements and their explanations, detailed information on usage and sample input. (author)

  1. Corrosion Resistant Cladding by YAG Laser Welding in Underwater Environment

    International Nuclear Information System (INIS)

    Tsutomi Kochi; Toshio Kojima; Suemi Hirata; Ichiro Morita; Katsura Ohwaki

    2002-01-01

    It is known that stress-corrosion cracking (SCC) will occur in nickel-base alloys used in Reactor Pressure Vessel (RPV) and Internals of nuclear power plants. A SCC sensitivity has been evaluated by IHI in each part of RPV and Internals. There are several water level instrumentation nozzles installed in domestic BWR RPV. In water level instrumentation nozzles, 182 type nickel-base alloys were used for the welding joint to RPV. It is estimated the SCC potential is high in this joint because of a higher residual stress than the yield strength (about 400 MPa). This report will describe a preventive maintenance method to these nozzles Heat Affected Zone (HAZ) and welds by a corrosion resistant cladding (CRC) by YAG Laser in underwater environment (without draining a reactor water). There are many kinds of countermeasures for SCC, for example, Induction Heating Stress Improvement (IHSI), Mechanical Stress Improvement Process (MSIP) and so on. A YAG laser CRC is one of them. In this technology a laser beam is used for heat source and irradiated through an optical fiber to a base metal and SCC resistant material is used for welding wires. After cladding the HAZ and welds are coated by the corrosion resistant materials so their surfaces are improved. A CRC by gas tungsten arc welding (GTAW) in an air environment had been developed and already applied to a couple of operating plants (16 Nozzles). This method was of course good but it spent much time to perform because of an installation of some water-proof working boxes to make a TIG-weldability environment. CRC by YAG laser welding in underwater environment has superior features comparing to this conventional TIG method as follows. At the viewpoint of underwater environment, (1) an outage term reduction (no drainage water). (2) a radioactive exposure dose reduction for personnel. At that of YAG laser welding, (1) A narrower HAZ. (2) A smaller distortion. (3) A few cladding layers. A YAG laser CRC test in underwater

  2. Crud deposition modeling on BWR fuel rods

    International Nuclear Information System (INIS)

    Kucuk, Aylin; Cheng, Bo; Potts, Gerald A.; Shiralkar, Bharat; Morgan, Dave; Epperson, Kenny; Gose, Garry

    2014-01-01

    Deposition of boiling water reactor (BWR) system corrosion products (crud) on operating fuel rods has resulted in performance-limiting conditions in a number of plants. The operational impact of performance-limiting conditions involving crud deposition can be detrimental to a BWR operator, resulting in unplanned or increased frequency of fuel inspections, fuel failure and associated radiological consequences, operational restrictions including core power derate and/or forced shutdowns to remove failed fuel, premature discharge of individual bundles or entire reloads, and/or undesirable core design restrictions. To facilitate improved management of crud-related fuel performance risks, EPRI has developed the CORAL (Crud DepOsition Risk Assessment ModeL) tool. This paper presents a summary of the CORAL elements and benchmarking results. Applications of CORAL as a tool for fuel performance risk assessment are also discussed. (author)

  3. BWR radiation buildup control with ionic zinc

    International Nuclear Information System (INIS)

    Marble, W.J.; Wood, C.J.; Leighty, C.E.; Green, T.A.

    1986-01-01

    In 1983 a hypothesis was disclosed which suggested that the presence of ionic zinc in the reactor water of the BWR could reduce radiation buildup. This hypothesis was developed from correlations of plant data, and subsequently, from laboratory experiments which demonstrated clearly that ionic zinc inhibits the corrosion of stainless steel. The benefits of zinc addition have been measured at the Vallecitos Nuclear Center under and EPRI/GE project. Experimentation and analyses have been performed to evaluate the impact of intentional zinc addition on the IGSCC characteristics of primary system materials and on the performance of the nuclear fuel. It has been concluded that no negative effects are expected. The author conclude that the intentional addition of ionic zinc to the BWR reactor water at a concentration of approximately 10 ppb will provide major benefits in controlling the Co-60 buildup on primary system stainless steel surfaces. The intentional addition of zinc is now a qualified technique for use in BWRs

  4. BWR mechanics and materials technology update

    International Nuclear Information System (INIS)

    Kiss, E.

    1983-01-01

    This paper discusses technical results obtained from a variety of important programs underway at General Electric's Nuclear Engineering Division. The principal objective of these programs is to qualify and improve BWR product related technologies that fall broadly under the disciplines of Applied Mechanics and Materials Engineering. The paper identifies and deals with current technical issues that are of general importance to the LWR industry albeit the specific focus is directed to the development and qualification of analytical predictive methods and criteria, and improved materials for use in the design of the BWR. In this paper, specific results and accomplishments are summarized to provide a braod perspective of technology advances. Results are presented in sections which discuss: dynamic analysis and modeling; fatigue and fracture evaluation; materials engineering advances; and flow induced vibration. (orig.)

  5. Interaction between thorium and potential clad materials

    International Nuclear Information System (INIS)

    Kale, G.B.; Gawde, P.S.; Sengupta, Pranesh

    2005-01-01

    Thorium based fuels are being used for nuclear reactors. The structural stability of fuel-clad assemblies in reactor systems depend upon the nature of interdiffusion reaction between fuel-cladding materials. Interdiffusion reaction thorium and various cladding materials is presented in this paper. (author)

  6. Eulerian fluid-structure analysis of BWR

    International Nuclear Information System (INIS)

    McMaster, W.H.

    1979-05-01

    A fluid-structure-interaction algorithm is developed for the analysis of the dynamic response of a BWR pressure-suppression pool and containment structure. The method is incorporated into a two-dimensional semi-implicit Eulerian hydrodynamics code, PELE-IC, for the solution of incompressible flow coupled to flexible structures. The fluid, structure, and coupling algorithms have been verified by calculation of solved problems from the literature and by comparison with air and steam blowdown experiments

  7. Methyl Iodide Decomposition at BWR Conditions

    International Nuclear Information System (INIS)

    Pop, Mike; Bell, Merl

    2012-09-01

    Based on favourable results from short-term testing of methanol addition to an operating BWR plant, AREVA has performed numerous studies in support of necessary Engineering and Plant Safety Evaluations prior to extended injection of methanol. The current paper presents data from a study intended to provide further understanding of the decomposition of methyl iodide as it affects the assessment of methyl iodide formation with the application of methanol at BWR Plants. This paper describes the results of the decomposition testing under UV-C light at laboratory conditions and its effect on the subject methyl iodide production evaluation. The study as to the formation and decomposition of methyl iodide as it is effected by methanol addition is one phase of a larger AREVA effort to provide a generic plant Safety Evaluation prior to long-term methanol injection to an operating BWR. Other testing phases have investigated the compatibility of methanol with fuel construction materials, plant structural materials, plant consumable materials (i.e. elastomers and coatings), and ion exchange resins. Methyl iodide is known to be very unstable, typically preserved with copper metal or other stabilizing materials when produced and stored. It is even more unstable when exposed to light, heat, radiation, and water. Additionally, it is known that methyl iodide will decompose radiolytically, and that this effect may be simulated using ultra-violet radiation (UV-C) [2]. In the tests described in this paper, the use of a UV-C light source provides activation energy for the formation of methyl iodide. Thus is similar to the effect expected from Cherenkov radiation present in a reactor core after shutdown. Based on the testing described in this paper, it is concluded that injection of methanol at concentrations below 2.5 ppm in BWR applications to mitigate IGSCC of internals is inconsequential to the accident conditions postulated in the FSAR as they are related to methyl iodide formation

  8. Delivering high performance BWR fuel reliably

    Energy Technology Data Exchange (ETDEWEB)

    Schardt, J.F. [GE Nuclear Energy, Wilmington, NC (United States)

    1998-07-01

    Utilities are under intense pressure to reduce their production costs in order to compete in the increasingly deregulated marketplace. They need fuel, which can deliver high performance to meet demanding operating strategies. GE's latest BWR fuel design, GE14, provides that high performance capability. GE's product introduction process assures that this performance will be delivered reliably, with little risk to the utility. (author)

  9. The HAMBO BWR simulator of HAMMLAB

    International Nuclear Information System (INIS)

    Karlsson, Tommy; Jokstad, Haakon; Meyer, Brita D.; Nihlwing, Christer; Norrman, Sixten; Puska, Eija Karita; Raussi, Pekka; Tiihonen, Olli

    2001-02-01

    Modernisation of control rooms of the nuclear power plants has been a major issue in Sweden and Finland the last few years, and this will continue in the years to come. As an aid in the process of introducing new technology into the control rooms, the benefit of having an experimental simulator where proto typing of solutions can be performed, has been emphasised by many plants. With this as a basis, the BWR plants in Sweden and Finland decided to fund, in co-operation with the Halden Project, an experimental BWR simulator based on the Forsmark 3 plant in Sweden. The BWR simulator development project was initiated in January 1998. VTT Energy in Finland developed the simulator models with the aid of their APROS tool, while the operator interface was developed by the Halden Project. The simulator was thoroughly tested by experienced HRP personnel and professional Forsmark 3 operators, and accepted by the BWR utilities in June 2000. The acceptance tests consisted of 19 well-defined transients, as well as the running of the simulator from full power down to cold shutdown and back up again with the use of plant procedures. This report describes the HAMBO simulator, with its simulator models, the operator interface, and the underlying hardware and software infrastructure. The tools used for developing the simulator, APROS, Picasso-3 and the Integration Platform, are also briefly described. The acceptance tests are described, and examples of the results are presented, to illustrate the level of validation of the simulator. The report concludes with an indication of the short-term usage of the simulator. (Author)

  10. PREDICTIVE METHODS FOR STABILITY MARGIN IN BWR

    OpenAIRE

    MELARA SAN ROMÁN, JOSÉ

    2016-01-01

    [EN] Power and flow oscillations in a BWR are very undesirable. One of the major concerns is to ensure, during power oscillations, compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions ...

  11. BWR level estimation using Kalman Filtering approach

    International Nuclear Information System (INIS)

    Garner, G.; Divakaruni, S.M.; Meyer, J.E.

    1986-01-01

    Work is in progress on development of a system for Boiling Water Reactor (BWR) vessel level validation and failure detection. The levels validated include the liquid level both inside and outside the core shroud. This work is a major part of a larger effort to develop a complete system for BWR signal validation. The demonstration plant is the Oyster Creek BWR. Liquid level inside the core shroud is not directly measured during full power operation. This level must be validated using measurements of other quantities and analytic models. Given the available sensors, analytic models for level that are based on mass and energy balances can contain open integrators. When such a model is driven by noisy measurements, the model predicted level will deviate from the true level over time. To validate the level properly and to avoid false alarms, the open integrator must be stabilized. In addition, plant parameters will change slowly with time. The respective model must either account for these plant changes or be insensitive to them to avoid false alarms and maintain sensitivity to true failures of level instrumentation. Problems are addressed here by combining the extended Kalman Filter and Parity Space Decision/Estimator. The open integrator is stabilized by integrating from the validated estimate at the beginning of each sampling interval, rather than from the model predicted value. The model is adapted to slow plant/sensor changes by updating model parameters on-line

  12. Utility experience with BWR-PSMS

    International Nuclear Information System (INIS)

    Bond, G.R.

    1986-01-01

    The BWR Power Shape Monitoring System (BWR-PSMS) has proven to be an effective and versatile tool for core monitoring. GPU Nuclear Corporation's (GPUN) Oyster Creek plant has been involved in the PSMS development since its inception, having been selected by EPRI as the initial demonstration site. Beginning with Cycle 10, Oyster Creek has been applying the BWR-PSMS as the primary core monitoring tool. Although the system has been in operation at Oyster Creek for the past several cycles, this is the first time the PSMS was used to monitor compliance to the plant technical specifications, to guide adherence to vendore fuel maneuvering recommendations and to develop data for certain performance records such as fuel burnup, isotopic accounting, etc. This paper will discuss the bases for the decision to apply PSMS as the fundamental core monitoring system, the experience in implementing the PSMS in this mode, activities currently underway or planned related to PSMS, and potential future extensions and applications of PSMS at Oyster Creek

  13. Economic analysis of hydride fueled BWR

    International Nuclear Information System (INIS)

    Ganda, F.; Shuffler, C.; Greenspan, E.; Todreas, N.

    2009-01-01

    The economic implications of designing BWR cores with hydride fuels instead of conventional oxide fuels are analyzed. The economic analysis methodology adopted is based on the lifetime levelized cost of electricity (COE). Bracketing values (1970 and 3010 $/kWe) are used for the overnight construction costs and for the power scaling factors (0.4 and 0.8) that correlate between a change in the capital cost to a change in the power level. It is concluded that a newly constructed BWR reactor could substantially benefit from the use of 10 x 10 hydride fuel bundles instead of 10 x 10 oxide fuel bundles design presently in use. The cost saving would depend on the core pressure drop constraint that can be implemented in newly constructed BWRs - it is between 2% and 3% for a core pressure drop constraint as of the reference BWR, between 9% and 15% for a 50% higher core pressure drop, and between 12% and 21% higher for close to 100% core pressure. The attainable cost reduction was found insensitive to the specific construction cost but strongly dependent on the power scaling factor. The cost advantage of hydride fuelled cores as compared to that of the oxide reference core depends only weakly on the uranium and SWU prices, on the 'per volume base' fabrication cost of hydride fuels, and on the discount rate used. To be economically competitive, the uranium enrichment required for the hydride fuelled core needs to be around 10%.

  14. Investigation of BWR stability in Forsmark 2

    International Nuclear Information System (INIS)

    Oguma, R.; Reisch, F.; Bergdahl, B.G.; Lorenzen, J.; Aakerhielm, F.; Kellner, S.

    1988-01-01

    A series of noise measurements have been conducted at the Forsmark-2 reactor during its start-up operation after the revision in 1987. The main purpose was to investigate the BWR stability problem based on noise analysis, i.e. the problem of resonant power oscillation with frequency of about 0.5 Hz, which tends to arise at high power and low core flow condition. The noise analysis was performed to estimate the noise source which gives rise to the power oscillation, to evaluate the stability condition of the Forsmark-2 reactor in terms of the decay ratio (DR), as well as to investigate a safety related problem in connection with the BWR stability. The results indicate that the power oscillation is due to dynamic coupling between the neutron kinetics and thermal-hydraulics via void reactivity feedback. The DR reached as high as ≅ 0.7 at 63% of the rated power and 4100 kg/s of the total core flow. An investigation was made for the noise recording which represents a strong pressure oscillation with a peak frequency at 0.33 Hz. The result suggests that such pressure oscillation, if the peak frequency coincided with that of the resonant power oscillation, might become a cause of scram. The present noise analysis indicates the importance of a BWR on-line surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  15. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1994-01-01

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

  16. Scaling and uncertainty in BWR instability problems

    International Nuclear Information System (INIS)

    Di Auria, F.; Pellicoro, V.

    1995-01-01

    This paper deals with a critical review of activities, performed at the DCMN of Pisa University, in relation to the thermo-hydraulic oscillations in two-phase systems. Stability analyses, including model development and achievement of experimental data, are generally performed for BWRs in order to achieve the following objectives: to reach a common understanding in relation to the predictive capabilities of system codes and to the influence of various parameters on the instability; to establish a data base for the qualification of the analytical tools already or becoming available; to set-up qualified tools (code/models + nodalization + user assumption) suitable for predicting the unstable behaviour of the nuclear plants of interest (current BWR, SBWR, ABWR and RBMK). These considerations have been the basis for the following researches: 1) proposal of the Boiling Instability Program (BIP) (1) 2) evaluation of stability tests in PIPER-ONE apparatus (2) 3) coupled thermal-hydraulic and neutronic instabilities in the LaSalle-2 BWR plant (3) 4) participation to the NEA-OECD BWR Benchmark (4) The RELAP/MOD2 and RELAP5/MOD3 codes have been used. (author)

  17. BWR recirculation pump diagnostic expert system

    International Nuclear Information System (INIS)

    Chiang, S.C.; Morimoto, C.N.; Torres, M.R.

    2004-01-01

    At General Electric (GE), an on-line expert system to support maintenance decisions for BWR recirculation pumps for nuclear power plants has been developed. This diagnostic expert system is an interactive on-line system that furnishes diagnostic information concerning BWR recirculation pump operational problems. It effectively provides the recirculation pump diagnostic expertise in the plant control room continuously 24 hours a day. The expert system is interfaced to an on-line monitoring system, which uses existing plant sensors to acquire non-safety related data in real time. The expert system correlates and evaluates process data and vibration data by applying expert rules to determine the condition of a BWR recirculation pump system by applying knowledge based rules. Any diagnosis will be automatically displayed, indicating which pump may have a problem, the category of the problem, and the degree of concern expressed by the validity index and color hierarchy. The rules incorporate the expert knowledge from various technical sources such as plant experience, engineering principles, and published reports. These rules are installed in IF-THEN formats and the resulting truth values are also expressed in fuzzy terms and a certainty factor called a validity index. This GE Recirculation Pump Expert System uses industry-standard software, hardware, and network access to provide flexible interfaces with other possible data acquisition systems. Gensym G2 Real-Time Expert System is used for the expert shell and provides the graphical user interface, knowledge base, and inference engine capabilities. (author)

  18. Interpretation of incore noise measurements in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van

    1982-01-01

    A survey is given of the main incentives for power reactor noise research and the differences and similarities of noise in power and zero power systems are touched on. The basic characteristics of the adjoint method in reactor noise theory are treated. The detector adjoint functions describe the transfer functions between spatially distributed noise sources and a (neutron or gamma) detector. In particular, the spatial dependence of these functions explains the 'local' and 'global' effects in BWR noise measurements. By including thermal hydraulic feedback effects in the adjoint analysis, it is shown that the common idea of a dominant global effect at low frequencies which should result in point kinetic behaviour, is erroneous. The same analysis provides a method for nonperturbing on-line measurement of the reactor transfer function, which is demonstrated by results from measurements on a BWR in the Netherlands. In the final part of the paper some ideas are given for further research in the field of BWR noise. (author)

  19. High fidelity analysis of BWR fuel assembly with COBRA-TF/PARCS and trace codes

    International Nuclear Information System (INIS)

    Abarca, A.; Miro, R.; Barrachina, T.; Verdu, G.; Soler, A.

    2013-01-01

    The growing importance of detailed reactor core and fuel assembly description for light water reactors (LWRs) as well as the sub-channel safety analysis requires high fidelity models and coupled neutronic/thermalhydraulic codes. Hand in hand with advances in the computer technology, the nuclear safety analysis is beginning to use a more detailed thermal hydraulics and neutronics. Previously, a PWR core and a 16 by 16 fuel assembly models were developed to test and validate our COBRA-TF/PARCS v2.7 (CTF/PARCS) coupled code. In this work, a comparison of the modeling and simulation advantages and disadvantages of modern 10 by 10 BWR fuel assembly with CTF/PARCS and TRACE codes has been done. The objective of the comparison is making known the main advantages of using the sub-channel codes to perform high resolution nuclear safety analysis. The sub-channel codes, like CTF, permits obtain accurate predictions, in two flow regime, of the thermalhydraulic parameters important to safety with high local resolution. The modeled BWR fuel assembly has 91 fuel rods (81 full length and 10 partial length fuel rods) and a big square central water rod. This assembly has been modeled with high level of detail with CTF code and using the BWR modeling parameters provided by TRACE. The same neutronic PARCS's model has been used for the simulation with both codes. To compare the codes a coupled steady state has be performed. (author)

  20. A statistical analysis of pellet-clad interaction failures in water reactor fuel

    International Nuclear Information System (INIS)

    McDonald, S.G.; Fardo, R.D.; Sipush, P.J.; Kaiser, R.S.

    1981-01-01

    The primary objective of the statistical analysis was to develop a mathematical function that would predict PCI fuel rod failures as a function of the imposed operating conditions. Linear discriminant analysis of data from both test and commercial reactors was performed. The initial data base used encompassed 713 data points (117 failures and 596 non-failures) representing a wide variety of water cooled reactor fuel (PWR, BWR, CANDU, and SGHWR). When applied on a best-estimate basis, the resulting function simultaneously predicts approximately 80 percent of both the failure and non-failure data correctly. One of the most significant predictions of the analysis is that relatively large changes in power can be tolerated when the pre-ramp irradiation power is low, but that only small changes in power can be tolerated when the pre-ramp irradiation power is high. However, it is also predicted that fuel rods irradiated at low power will fail at lower final powers than those irradiated at high powers. Other results of the analysis are that fuel rods with high clad operating temperatures can withstand larger power increases that fuel rods with low clad operating temperatures, and that burnup has only a minimal effect on PCI performance after levels of approximately 10000 MWD/MTU have been exceeded. These trends in PCI performance and the operating parameters selected are believed to be consistent with mechanistic considerations. Published PCI data indicate that BWR fuel usually operates at higher local powers and changes in power, lower clad temperatures, and higher local ramp rates than PWR fuel

  1. ZIRCONIUM-CLADDING OF THORIUM

    Science.gov (United States)

    Beaver, R.J.

    1961-11-21

    A method of cladding thorium with zirconium is described. The quality of the bond achieved between thorium and zirconium by hot-rolling is improved by inserting and melting a thorium-zirconium alloy foil between the two materials prior to rolling. (AEC)

  2. Clad-coolant chemical interaction

    International Nuclear Information System (INIS)

    Iglesias, F.C.; Lewis, B.J.; Desgranges, C.; Toffolon, C.

    2015-01-01

    This paper provides an overview of the kinetics for zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. Low-temperature oxidation of zircaloy due to water-side corrosion is further described. (authors)

  3. Development of high performance cladding

    International Nuclear Information System (INIS)

    Kiuchi, Kiyoshi

    2003-01-01

    The developments of superior next-generation light water reactor are requested on the basis of general view points, such as improvement of safety, economics, reduction of radiation waste and effective utilization of plutonium, until 2030 year in which conventional reactor plants should be renovate. Improvements of stainless steel cladding for conventional high burn-up reactor to more than 100 GWd/t, developments of manufacturing technology for reduced moderation-light water reactor (RMWR) of breeding ratio beyond 1.0 and researches of water-materials interaction on super critical pressure-water cooled reactor are carried out in Japan Atomic Energy Research Institute. Stable austenite stainless steel has been selected for fuel element cladding of advanced boiling water reactor (ABWR). The austenite stain less has the superiority for anti-irradiation properties, corrosion resistance and mechanical strength. A hard spectrum of neutron energy up above 0.1 MeV takes place in core of the reduced moderation-light water reactor, as liquid metal-fast breeding reactor (LMFBR). High performance cladding for the RMWR fuel elements is required to get anti-irradiation properties, corrosion resistance and mechanical strength also. Slow strain rate test (SSRT) of SUS 304 and SUS 316 are carried out for studying stress corrosion cracking (SCC). Irradiation tests in LMFBR are intended to obtain irradiation data for damaged quantity of the cladding materials. (M. Suetake)

  4. BWR Refill-Reflood Program, Task 4.7 - model development: TRAC-BWR component models

    International Nuclear Information System (INIS)

    Cheung, Y.K.; Parameswaran, V.; Shaug, J.C.

    1983-09-01

    TRAC (Transient Reactor Analysis Code) is a computer code for best-estimate analysis for the thermal hydraulic conditions in a reactor system. The development and assessment of the BWR component models developed under the Refill/Reflood Program that are necessary to structure a BWR-version of TRAC are described in this report. These component models are the jet pump, steam separator, steam dryer, two-phase level tracking model, and upper-plenum mixing model. These models have been implemented into TRAC-B02. Also a single-channel option has been developed for individual fuel-channel analysis following a system-response calculation

  5. Development of high performance cladding materials

    International Nuclear Information System (INIS)

    Park, Jeong Yong; Jeong, Y. H.; Park, S. Y.

    2010-04-01

    The irradiation test for HANA claddings conducted and a series of evaluation for next-HANA claddings as well as their in-pile and out-of pile performances tests were also carried out at Halden research reactor. The 6th irradiation test have been completed successfully in Halden research reactor. As a result, HANA claddings showed high performance, such as corrosion resistance increased by 40% compared to Zircaloy-4. The high performance of HANA claddings in Halden test has enabled lead test rod program as the first step of the commercialization of HANA claddings. DB has been established for thermal and LOCA-related properties. It was confirmed from the thermal shock test that the integrity of HANA claddings was maintained in more expanded region than the criteria regulated by NRC. The manufacturing process of strips was established in order to apply HANA alloys, which were originally developed for the claddings, to the spacer grids. 250 kinds of model alloys for the next-generation claddings were designed and manufactured over 4 times and used to select the preliminary candidate alloys for the next-generation claddings. The selected candidate alloys showed 50% better corrosion resistance and 20% improved high temperature oxidation resistance compared to the foreign advanced claddings. We established the manufacturing condition controlling the performance of the dual-cooled claddings by changing the reduction rate in the cold working steps

  6. Study of laser cladding nuclear valve parts

    International Nuclear Information System (INIS)

    Shi Shihong; Wang Xinlin; Huang Guodong

    1998-12-01

    The mechanism of laser cladding is discussed by using heat transfer model of laser cladding, heat conduction model of laser cladding and convective transfer mass model of laser melt-pool. Subsequently the laser cladding speed limit and the influence of laser cladding parameters on cladding layer structure is analyzed. A 5 kW with CO 2 transverse flow is used in the research for cladding treatment of sealing surface of stop valve parts of nuclear power stations. The laser cladding layer is found to be 3.0 mm thick. The cladding surface is smooth and has no such defects as crack, gas pore, etc. A series of comparisons with plasma spurt welding and arc bead welding has been performed. The results show that there are higher grain grade and hardness, lower dilution and better performances of resistance to abrasion, wear and of anti-erosion in the laser cladding layer. The new technology of laser cladding can obviously improve the quality of nuclear valve parts. Consequently it is possible to lengthen the service life of nuclear valve and to raise the safety and reliability of the production system

  7. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  8. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  9. GPE-BWR and the containment venting and filtering issue

    International Nuclear Information System (INIS)

    Palomo, J.; Santiago, J. de

    1988-01-01

    The Spanish Boiling Water Reactor Owner's Group (GPE-BWR) is formed by three utilities, owning four units: Santa Maria de Garona (46 MWe, BWR3, Mark I containment), Cofrentes (975 MWe, BWR6, Mark III containment) and Valdecaballeros (2x975 MWe, BWR6, Mark III containment) - all of the reactors having been supplied by General Electric. One of the GPE-BWR's several committees is the Safety and Licensing Committee, which follows up the evolution of severe accident topics and particularly the containment venting and filtering issue. In September 1987, the Consejo de Seguridad Nuclear (CSN), the Spanish Regulatory Body, asked the GPE-BWR to define its position on the installation of a containment venting system. The GPE-BWR created a Working Group which presented a Report on Containment Venting to the CSN in January 1987 gathered from: the US Nuclear Regulatory Commission (NRC); some US utilities; and several European countries, especially France, Germany and Sweden. CSN's review of the containment venting Report and the Action Plan proposed by the GPE-BWR finished in April 1988. The conclusion of the Report and the proposed Action Plan take into account the US NRC's identified open items on severe accidents and the R and D programs scheduled to close these items

  10. BWR Full Integral Simulation Test (FIST) Phase II test results and TRAC-BWR model qualification

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Findlay, J.A.; Hwang, W.S.

    1985-10-01

    Eight matrix tests were conducted in the FIST Phase I. These tests investigated the large break, small break and steamline break LOCA's, as well as natural circulation and power transients. There are nine tests in Phase II of the FIST program. They include the following LOCA tests: BWR/6 LPCI line break, BWR/6 intermediate size recirculation break, and a BWR/4 large break. Steady state natural circulation tests with feedwater makeup performed at high and low pressure, and at high pressure with HPCS makeup, are included. Simulation of a transient without rod insertion, and with controlled depressurization, was performed. Also included is a simulation of the Peach Bottom turbine trip test. The final two tests simulated a failure to maintain water level during a postulated accident. A FIST program objective is to assess the TRAC code by comparisons with test data. Two post-test predictions made with TRACB04 are compared with Phase II test data in this report. These are for the BWR/6 LPCI line break LOCA, and the Peach Bottom turbine trip test simulation

  11. Laser cladding of quasicrystalline alloys

    International Nuclear Information System (INIS)

    Audebert, F.; Sirkin, H.; Colaco, R.; Vilar, R.

    1998-01-01

    Quasicrystals are a new class of ordinated structures with metastable characteristics room temperature. Quasicrystalline phases can be obtained by rapid quenching from the melt of some alloys. In general, quasicrystals present properties which make these alloys promising for wear and corrosion resistant coatings applications. During the last years, the development of quasicrystalline coatings by means of thermal spray techniques has been impulsed. However, no references have been found of their application by means of laser techniques. In this work four claddings of quasicrystalline compositions formed over aluminium substrate, produced by a continuous CO 2 laser using simultaneous powders mixture injection are presented. The claddings were characterized by X ray diffraction, scanning electron microscopy and Vickers microhardness. (Author) 18 refs

  12. Prevention of organic iodide formation in BWR's

    International Nuclear Information System (INIS)

    Karjunen, T.; Laitinen, T.; Piippo, J.; Sirkiae, P.

    1996-01-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR's as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs

  13. CLAD DEGRADATION - FEPS SCREENING ARGUMENTS

    International Nuclear Information System (INIS)

    R. Schreiner

    2004-01-01

    The purpose of this report is to evaluate and document the screening of the clad degradation features, events, and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment-License Application (TSPA-LA). This report also addresses the effect of certain FEPs on both the cladding and the commercial spent nuclear fuel (CSNF), DOE-owned spent nuclear fuel (DSNF), and defense high-level waste (DHLW) waste forms, as appropriate to address the effects on multiple materials and both components (FEPs 2.1.09.09.0A, 2.1.09.11.0A, 2.1.11.05.0A, 2.1.12.02.0A, and 2.1.12.03.0A). These FEPs are expected to affect the repository performance during the postclosure regulatory period of 10,000 years after permanent closure. Table 1-1 provides the list of cladding FEPs, including their screening decisions (include or exclude). The primary purpose of this report is to identify and document the analysis, screening decision, and TSPA-LA disposition (for included FEPs) or screening argument (for excluded FEPs) for these FEPs related to clad degradation. In some cases, where a FEP covers multiple technical areas and is shared with other FEP reports, this report may provide only a partial technical basis for the screening of the FEP. The full technical basis for shared FEPs is addressed collectively by the sharing FEP reports. The screening decisions and associated TSPA-LA dispositions or screening arguments from all of the FEP reports are cataloged in a project-specific FEPs database

  14. A simplified spatial model for BWR stability

    International Nuclear Information System (INIS)

    Berman, Y.; Lederer, Y.; Meron, E.

    2012-01-01

    A spatial reduced order model for the study of BWR stability, based on the phenomenological model of March-Leuba et al., is presented. As one dimensional spatial dependence of the neutron flux, fuel temperature and void fraction is introduced, it is possible to describe both global and regional oscillations of the reactor power. Both linear stability analysis and numerical analysis were applied in order to describe the parameters which govern the model stability. The results were found qualitatively similar to past results. Doppler reactivity feedback was found essential for the explanation of the different regions of the flow-power stability map. (authors)

  15. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

    2007-01-01

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

  16. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  17. Maintenance of BWR control rod drive mechanisms

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    Control rod drive mechanism (CRDM) replacement and rebuilding is one of the highest dose, most physically demanding, and complicated maintenance activities routinely accomplished by BWR utilities. A recent industry workshop sponsored by the Oak Ridge National Laboratory, which dealt with the effects of CRDM aging, revealed enhancements in maintenance techniques and tooling which have reduced ALARA, improved worker comfort and productivity, and have provided revised guidelines for CRDM changeout selection. Highlights of this workshop and ongoing research on CRDM aging are presented in this paper

  18. Kinetic analyses on startup and shutdown chemistry of BWR plant

    International Nuclear Information System (INIS)

    Domae, Masafumi; Fujiwara, Kazutoshi; Inagaki, Hiromitsu

    2012-09-01

    During startup and shutdown of Boiling Water Reactor (BWR) plants, temperature and dissolved oxygen (DO) concentration of reactor water change in a wide range. The changes result in variation of conductivity and pH of the reactor water. It has been speculated that the water chemistry change is due to dissolution of the oxides on fuel claddings and structural materials. However, detailed mechanism is not known. In the present paper, trend of recent water chemistry in several BWR plants during startup and shutdown is presented. Conductivity and pH are convenient indication of coolant purity. We tried to clarify the mechanism of the change in the conductivity and the pH value during startup and shutdown, based on the water chemistry data measured. In the water chemistry data, change in chromate concentration and Ni 2+ concentration is rather large. It is assumed that change in the chromate concentration and the Ni 2+ concentration results in the time variation of the conductivity and the pH value. It is reasonable to consider that the increase in the chromate concentration and the Ni 2+ concentration is ascribed to dissolution of Cr oxides and Ni oxides, respectively. A model of dissolution of the Cr oxides and the Ni oxides is proposed. A concept of finite inventory of the Cr oxides and the Ni oxides in the coolant system is introduced. The model is as follows. Chromate is generated by oxidation of the Cr oxides and the Cr dissolution rate depends on the DO concentration. The dissolution rate of chromate is in proportion to DO concentration, the inventory of Cr and difference between solubility limit and the chromate concentration. On the other hand, Ni 2+ is formed by dissolution of the Ni oxides, and DO is not necessary in this process. The dissolution rate of Ni 2+ is in proportion to the inventory of Ni and difference between solubility limit and the Ni 2+ concentration. Coolant is continuously purified, and the chromate concentration and the Ni 2+ concentration

  19. Review and evaluation of cladding attack of LMFBR fuel

    International Nuclear Information System (INIS)

    Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T.

    1977-01-01

    The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio (oxygen to metal ratio), burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature

  20. Fundamentals and industrial applications of high power laser beam cladding

    International Nuclear Information System (INIS)

    Bruck, G.J.

    1988-01-01

    Laser beam cladding has been refined such that clad characteristics are precisely determined through routine process control. This paper reviews the state of the art of laser cladding optical equipment, as well as the fundamental process/clad relationships that have been developed for high power processing. Major categories of industrial laser cladding are described with examples chose to highlight particular process attributes

  1. Unirradiated cladding rip-propagation tests

    International Nuclear Information System (INIS)

    Hu, W.L.; Hunter, C.W.

    1981-04-01

    The size of cladding rips which develop when a fuel pin fails can affect the subassembly cooling and determine how rapidly fuel escapes from the pin. The object of the Cladding Rip Propagation Test (CRPT) was to quantify the failure development of cladding so that a more realistic fuel pin failure modeling may be performed. The test results for unirradiated 20% CS 316 stainless steel cladding show significantly different rip propagation behavior at different temperatures. At room temperature, the rip growth is stable as the rip extension increases monotonically with the applied deformation. At 500 0 C, the rip propagation becomes unstable after a short period of stable rip propagation. The rapid propagation rate is approximately 200 m/s, and the critical rip length is 9 mm. At test temperatures above 850 0 C, the cladding exhibits very high failure resistances, and failure occurs by multiple cracking at high cladding deformation. 13 figures

  2. Effect of laser power on clad metal in laser-TIG combined metal cladding

    Science.gov (United States)

    Utsumi, Akihiro; Hino, Takanori; Matsuda, Jun; Tasoda, Takashi; Yoneda, Masafumi; Katsumura, Munehide; Yano, Tetsuo; Araki, Takao

    2003-03-01

    TIG arc welding has been used to date as a method for clad welding of white metal as bearing material. We propose a new clad welding process that combines a CO2 laser and a TIG arc, as a method for cladding at high speed. We hypothesized that this method would permit appropriate control of the melted quantity of base metal by varying the laser power. We carried out cladding while varying the laser power, and investigated the structure near the boundary between the clad layer and the base metal. Using the laser-TIG combined cladding, we found we were able to control appropriately the degree of dilution with the base metal. By applying this result to subsequent cladding, we were able to obtain a clad layer of high quality, which was slightly diluted with the base metal.

  3. Mobile crud and transportation of radioactivity in BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hermansson, H-P. [Studsvik Nuclear AB, Nykoping (Sweden); LTU, Div. of Chemical Engineering, Lulea (Sweden); Hagg, J. [Ringhals AB, Varobacka (Sweden)

    2010-07-01

    Mobile crud is here referred to as a generic term for all types of particles that occur in the reactor water in BWRs and that are able to carry radioactivity. Previous results in this on-going series of studies in Swedish BWRs suggest that there are particles of different origins and function. A share may come from fuel crud and others may come from detachment, precipitation and dissolution processes in different parts of the BWR primary system, as well as from other system parts, such as the turbine/condenser. In addition, crud particles in this sense may come from purely mechanical processes such as degradation of graphite containing parts of the control rod drives. Therefore, the overall aim was to evaluate which particles are responsible for the transportation and distribution of radioactivity and also to clarify the chemical conditions under which they are formed. Furthermore the aim was to draw conclusions about how the chemistry would be like in order to avoid or at least minimize the formation of radioactivity distributing particles. A specific objective has also been to look into the importance of particle size for spreading of radioactivity in the primary system. Different types of crud particles are likely to have different characteristics in terms of function associated with transportation of radioactivity. The fuel crud is radioactive from the source and other types of crud can via surface processes, co-precipitation and other chemical and mechanical processes potentially affect the distribution of radioactivity in the primary system. In order to predict how operating parameters (e.g. stable, full power operation and scram) and chemical parameters (NWC/HWC/Zn, etc.) will affect the activity build-up on the system surfaces, it is important to know how the different types of crud are affected by these and related parameters. Fuel crud fixed on cladding ring samples, as well as mobile crud from the reactor water captured on filters, were examined by

  4. LASER SURFACE CLADDING FOR STRUCTURAL REPAIR

    OpenAIRE

    SANTANU PAUL

    2018-01-01

    Laser cladding is a powder deposition technique, which is used to deposit layers of clad material on a substrate to improve its surface properties. It has widespread application in the repair of dies and molds used in the automobile industry. These molds and dies are subjected to cyclic thermo-mechanical loading and therefore undergo localized damage and wear. The final clad quality and integrity is influenced by various physical phenomena, namely, melt pool morphology, microst...

  5. Phenomenology of BWR fuel assembly degradation

    Science.gov (United States)

    Kurata, Masaki; Barrachin, Marc; Haste, Tim; Steinbrueck, Martin

    2018-03-01

    Severe accidents occurred at the Fukushima-Daiichi Nuclear Power Station (FDNPS) which required an immediate re-examination of fuel degradation phenomenology. The present paper reviews the updated knowledge on the phenomenology of the fuel degradation, focusing mainly on the BWR fuel assembly degradation at the macroscopic scale and that of the individual interactions at the meso-scale. Oxidation of boron carbide (B4C) control rods potentially generates far larger amounts of heat and hydrogen under BWR accident conditions. All integral tests with B4C control rods or control blades have shown early failure, liquefaction, relocation and oxidation of B4C starting at temperatures around 1250 °C, well below the significant interaction temperatures of UO2-Zry. These interactions or reactions potentially influence the progress of fuel degradation in the early phase. The steam-starved conditions, which are being discussed as a likely scenario at the FDNPS accident, highly influence the individual interactions and potentially lead the fuel degradation in non-prototypical directions. The detailed phenomenology of individual interactions and their influence on the transient and on the late phase of the severe accidents are also discussed.

  6. BWR normal water chemistry guidelines: 1986 revision

    International Nuclear Information System (INIS)

    1988-09-01

    Boiling water reactors (BWRs) have experienced stress corrosion cracking in the reactor cooling system piping resulting in adverse impacts on plant availability and personnel radiation exposure. The BWR Owners Group and EPRI have sponsored a major research and development program to provide remedies for this stress corrosion cracking problem. This work shows that the likelihood of cracking depends on the plant's water chemistry performance (environment) as well as on material condition and stress level. Plant experience and other research demonstrate that water quality also affects fuel performance and radiation field buildup in BWRs. This report,''BWR Normal Water Chemistry Guidelines: 1986 Revision,'' presents suggested generic water chemistry specifications, justifies the proposed water chemistry limits, suggests responses to out-of-specification water chemistry, discusses available chemical analysis methods as well as data management and surveillance schemes, and details the management philosophy required to successfully implement a water chemistry control program. An appendix contains recommendations for water quality of auxiliary systems. 73 refs., 20 figs., 9 tabs

  7. BWR Assembly Optimization for Minor Actinide Recycling

    International Nuclear Information System (INIS)

    Maldonado, G. Ivan; Christenson, John M.; Renier, J.P.; Marcille, T.F.; Casal, J.

    2010-01-01

    The Primary objective of the proposed project is to apply and extend the latest advancements in LWR fuel management optimization to the design of advanced boiling water reactor (BWR) fuel assemblies specifically for the recycling of minor actinides (MAs). A top-level objective of the Advanced Fuel Cycle Systems Analysis program element of the DOE NERI program is to investigate spent fuel treatment and recycling options for current light water reactors (LWRs). Accordingly, this project targets to expand the traditional scope of nuclear fuel management optimization into the following two complementary specific objectives: (1) To develop a direct coupling between the pin-by-pin within-bundle loading control variables and core-wide (bundle-by-bundle) optimization objectives, (2) to extend the methodology developed to explicitly encompass control variables, objectives, and constraints designed to maximize minor actinide incineration in BWR bundles and cycles. The first specific objective is projected to 'uncover' dormant thermal margin made available by employing additional degrees of freedom within the optimization process, while the addition of minor actinides is expected to 'consume' some of the uncovered thermal margin. Therefore, a key underlying goal of this project is to effectively invest some of the uncovered thermal margin into achieving the primary objective.

  8. Recent BWR fuel management reactor physics advances

    International Nuclear Information System (INIS)

    Crowther, R.L.; Congdon, S.P.; Crawford, B.W.; Kang, C.M.; Martin, C.L.; Reese, A.P.; Savoia, P.J.; Specker, S.R.; Welchly, R.

    1982-01-01

    Improvements in BWR fuel management have been under development to reduce uranium and separative work (SWU) requirements and reduce fuel cycle costs, while also maintaining maximal capacity factors and high fuel reliability. Improved reactor physics methods are playing an increasingly important role in making such advances feasible. The improved design, process computer and analysis methods both increase knowledge of the thermal margins which are available to implement fuel management advance, and improve the capability to reliably and efficiently analyze and design for fuel management advances. Gamma scan measurements of the power distributions of advanced fuel assembly and advanced reactor core designs, and improved in-core instruments also are important contributors to improving 3-d predictive methods and to increasing thermal margins. This paper is an overview of the recent advances in BWR reactor physics fuel management methods, coupled with fuel management and core design advances. The reactor physics measurements which are required to confirm the predictions of performance fo fuel management advances also are summarized

  9. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  10. A BWR licensing experience in the USA

    International Nuclear Information System (INIS)

    Powers, J.; Ogura, C.; Arai, K.; Thomas, S.; Mookhoek, B.

    2015-09-01

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  11. BWR-stability investigation at Forsmark 1

    International Nuclear Information System (INIS)

    Bergdahl, B.G.; Reisch, F.; Oguma, R.; Lorenzen, J.; Aakerhielm, F.

    1988-01-01

    A series of noise measurements have been conducted at Forsmark 1 during start-up operation after the revision summer '87. The main purpose was to investigate BWR-stability problems, i.e. resonant power oscillations of 0.5 Hz around 65% power and 4100 kg/s core flow, which tend to arise at high power and low core flow conditions. The analysis was performed to estimate the noise source which gives rise to the oscillation, to evaluate the measure of stability, i.e. the Decay Ratio (Dr) as well as to investigate other safety related problems. The result indicates that the oscillation is due to the dynamic coupling between the neutron kinetics and thermal hydraulics via void reactivity feedback. The Dr ranged between values of 0.7 and > 0.9, instead of expected 0.6 (Dr=1 is defined as instability). These high values imply that the core cannot suppress oscillations fast enough and a small perturbation can cause scram. Further it was found that the entire core is oscillating in phase (LPRM's) with varying strength where any connection to the consequences of different fuel (8x8, 9x9) being present simultaneously cannot be excluded. This report elucidates the importance of an on-line BWR-stability surveillance system with functions like stability condition monitoring and control system diagnosis. (orig.)

  12. Hydrogen injection device in BWR type reactor

    International Nuclear Information System (INIS)

    Takagi, Jun-ichi; Kubo, Koji.

    1988-01-01

    Purpose: To reduce the increasing ratio of main steam system dose rate due to N-16 activity due to excess hydrogen injection in the hydrogen injection operation of BWR type reactors. Constitution: There are provided a hydrogen injection mechanism for injecting hydrogen into primary coolants of a BWR type reactor, and a chemical injection device for injecting chemicals such as methanol, which makes nitrogen radioisotopes resulted in the reactor water upon hydrogen injection non-volatile, into the pressure vessel separately from hydrogen. Injected hydrogen and the chemicals are not reacted in the feedwater system, but the reaction proceeds due to the presence of radioactive rays after the injection into the pressure vessel. Then, hydrogen causes re-combination in the downcomer portion to reduce the dissolved oxygen concentration. Meanwhile, about 70 % of the chemicals is supplied by means of a jet pump directly to the reactor core, thereby converting the chemical form of N-16 in the reactor core more oxidative (non-volatile). (Kawakami, Y.)

  13. A BWR licensing experience in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Powers, J.; Ogura, C. [Toshiba America Nuclear Energy, Charlotte, North Carolina (United States); Arai, K. [Toshiba Corporation, Yokohama, Kanagawa (Japan); Thomas, S.; Mookhoek, B., E-mail: jim.powers@toshiba.com [Nuclear Innovation North America, Lake Jackson, Texas (United States)

    2015-09-15

    The US-Advanced Boiling Water Reactor (A BWR), certified by the United States Nuclear Regulatory Commission (US NRC), is a third generation, evolutionary boiling water reactor design which is the reference for the South Texas Project Units 3 and 4 (STP3-4) Combined License Application (Cola). Nuclear Innovation North America (Nina) is the License Applicant for this new build project, and Toshiba is the selected primary technology contractor. The STP3-4 project has finished the US NRC technical review of the Cola through the final meeting of the Advisory Committee on Reactor Safeguards (ACRS), and the Final Safety Evaluation Report (FSER) is scheduled to be issued by the US NRC in the middle of 2015. The next steps are to support the Mandatory Hearing process, and voting by the NRC commissioners on the motion to grant the Combined License, which is scheduled beginning of 2016 according to US NRC schedule as of March 30, 2015. This paper summarizes the history and progress of the US-A BWR licensing, including the experiences of the Licensee, Nina, and Toshiba as the Epc team worked through the Code of Federal Regulations Title 10 (10-Cfr) Part 52 process, and provides some perspectives on how the related licensing material would also be of value within a 10-Cfr Part 50, two-step process to minimize schedule and financial risks which could arise from ongoing technical developments and regulatory reviews. (Author)

  14. BWR fuel experience with zinc injection

    International Nuclear Information System (INIS)

    Levin, H.A.; Garcia, S.E.

    1995-01-01

    In 1982 a correlation between low primary recirculation system dose rates in BWR's and the presence of ionic zinc in reactor water was identified. The source of the zinc was primarily from Admiralty brass condensers. Plants with brass condensers are called ''natural zinc'' plants. Brass condensers were also a source of copper that was implicated in crude induced localized corrosion (CILC) fuel failures. In 1986 the first BWR intentionally injected zinc for the benefits of dose rate control. Although zinc alone was never implicated in fuel degradation of failures, a comprehensive fuel surveillance program was initiated to monitor fuel performance. Currently there are 14 plants that are injecting zinc. Six of these plants are also on hydrogen water chemistry. This paper describes the effect on both Zircaloy corrosion and the cruding characteristics as a result of these changes in water chemistry. Fuel rod corrosion was found to be independent of the specific water chemistry of the plants. The corrosion behavior was the same with the additions of zinc alone or zinc plus hydrogen and well within the operating experience for fuel without either of these additions. No change was observed in the amounts of crude deposited on the fuel rods, both for the adherent and loosely held deposits. One of the effects of the zinc addition was the trend to form more of the zinc rich iron spinel in the fuel deposits rather than the hematite deposits that are predominantly formed with non additive water chemistry

  15. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-07-01

    The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon

  16. Laser surface cladding:a literature survey

    OpenAIRE

    Gedda, Hans

    2000-01-01

    This work consists of a literature survey of a laser surface cladding in order to investigate techniques to improve the cladding rate for the process. The high local heat input caused by the high power density of the laser generates stresses and the process is consider as slow when large areas are processed. To avoid these disadvantages the laser cladding process velocity can be increased three or four times by use of preheated wire instead of the powder delivery system. If laser cladding is ...

  17. Modelling cladding response to changing conditions

    Energy Technology Data Exchange (ETDEWEB)

    Tulkki, Ville; Ikonen, Timo [VTT Technical Research Centre of Finland ltd (Finland)

    2016-11-15

    The cladding of the nuclear fuel is subjected to varying conditions during fuel reactor life. Load drops and reversals can be modelled by taking cladding viscoelastic behaviour into account. Viscoelastic contribution to the deformation of metals is usually considered small enough to be ignored, and in many applications it merely contributes to the primary part of the creep curve. With nuclear fuel cladding the high temperature and irradiation as well as the need to analyse the variable load all emphasise the need to also inspect the viscoelasticity of the cladding.

  18. Pulsed Laser Cladding of Ni Based Powder

    Science.gov (United States)

    Pascu, A.; Stanciu, E. M.; Croitoru, C.; Roata, I. C.; Tierean, M. H.

    2017-06-01

    The aim of this paper is to optimize the operational parameters and quality of one step Metco Inconel 718 atomized powder laser cladded tracks, deposited on AISI 316 stainless steel substrate by means of a 1064 nm high power pulsed laser, together with a Precitec cladding head manipulated by a CLOOS 7 axes robot. The optimization of parameters and cladding quality has been assessed through Taguchi interaction matrix and graphical output. The study demonstrates that very good cladded layers with low dilution and increased mechanical proprieties could be fabricated using low laser energy density by involving a pulsed laser.

  19. Protective claddings for high strength chromium alloys

    Science.gov (United States)

    Collins, J. F.

    1971-01-01

    The application of a Cr-Y-Hf-Th alloy as a protective cladding for a high strength chromium alloy was investigated for its effectiveness in inhibiting nitrogen embrittlement of a core alloy. Cladding was accomplished by a combination of hot gas pressure bonding and roll cladding techniques. Based on bend DBTT, the cladding alloy was effective in inhibiting nitrogen embrittlement of the chromium core alloy for up to 720 ks (200hours) in air at 1422 K (2100 F). A significant increase in the bend DBTT occurred with longer time exposures at 1422 K or short time exposures at 1589 K (2400 F).

  20. Impact analysis of modifying the composition of the nuclear fuel of a BWR with beryllium oxide; Analisis del impacto de modificar la composicion del combustible nuclear de un BWR con oxido de berilio

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo V, J. M.; Morales S, J. B., E-mail: euqrop@hotmail.com [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    The beryllium oxide (Be O) presents excellent physical properties, especially its high thermal conductivity that contrasts clearly with that of the uranium dioxide (UO{sub 2}) used at the present as fuel in a great number of nuclear plants. The present work models a nuclear reactor cooled by light water in boiling with two external recirculation loops (BWR/5) using the code for the transitory analysis and postulated accidents Trac-B F1, implementing a UO{sub 2} mixture and different fractions of Be O, with the objective of improving the thermal conductivity of the fuel. The numeric results and the realized analyses indicate that when adding a fraction in volume of 10% the central temperature decreases in 30.4% in stationary state, while during the large break loss of coolant accident the peak cladding temperature diminishes in 7%. Although the real interaction of the mixture has not been determined experimentally, the obtained results are promising. (Author)

  1. Development of new irradiation facility for BWR safety research

    International Nuclear Information System (INIS)

    Okada, Yuji; Magome, Hirokatsu; Iida, Kazuhiro; Hanawa, Hiroshi; Ohmi, Masao

    2013-01-01

    In JAEA (Japan Atomic Energy Agency), about the irradiation embrittlement of the reactor pressure vessel and the stress corrosion cracking of reactor core composition apparatus concerning the long-term use of the light water reactor (BWR), in order to check the influence of the temperature, pressure, and water quality, etc on BWR condition. The water environmental control facility which performs irradiation assisted stress corrosion-cracking (IASCC) evaluation under BWR irradiation environment was fabricated in JMTR (Japan Materials Testing Reactor). This report is described the outline of manufacture of the water environmental control facility for doing an irradiation test using the saturation temperature capsule after JMTR re-operation. (author)

  2. Residual stress analysis in BWR pressure vessel attachments

    International Nuclear Information System (INIS)

    Dexter, R.J.; Leung, C.P.; Pont, D.

    1992-06-01

    Residual stresses from welding processes can be the primary driving force for stress corrosion cracking (SCC) in BWR components. Thus, a better understanding of the causes and nature of these residual stresses can help assess and remedy SCC. Numerical welding simulation software, such as SYSWELD, and material property data have been used to quantify residual stresses for application to SCC assessments in BWR components. Furthermore, parametric studies using SYSWELD have revealed which variables significantly affect predicted residual stress. Overall, numerical modeling techniques can be used to evaluate residual stress for SCC assessments of BWR components and to identify and plan future SCC research

  3. A pneumatic bellows-driven setup for controlled-distance electrochemical impedance measurements of Zircaloy-2 in simulated BWR conditions

    International Nuclear Information System (INIS)

    Arilahti, E.; Bojinov, M.; Hansson-Lyyra, L.

    2004-01-01

    This paper describes a novel pneumatic bellows-driven arrangement designed for controlled distance electrochemistry (CDE) measurements. The feasibility of the new arrangement has been verified by performing contact electric impedance measurements to study corrosion of Zircaloy-2 in a re-circulation loop simulating the BWR conditions. Until now, the measurements have been carried out using a step-motor driven controlled-distance electrochemistry (CDE) arrangement. The electrical and electrochemical properties of the pre transition oxide on Zircaloy-2 determined from these measurements were in good agreement with those estimated from measurements with a step-motor driven CDE. Furthermore, the results indicate that the bellows-driven CDE device is less sensitive to the contact pressure variation than the step-motor driven arrangement. This property combined with the bellows driven displacement mechanism provides a clear advantage for future in-core corrosion studies of fuel cladding materials. (Author)

  4. A PC Mathcad-based computational aid for severe accident analysis and its application to a BWR small LOCA sequence

    International Nuclear Information System (INIS)

    Wu, Laung-Kuang T.; Lee, S.J.

    2004-01-01

    A PC-based Mathcad program is used to develop a computational aid for analyzing severe accident phenomena. This computational aid uses simple engineering expressions and empirical correlations to estimate key quantities and timings at various stages of accident progressions. In this paper, the computational aid is applied to analyze an early phase of a BWR small LOCA sequence. The accident phenomena analyzed include: break flow rates, boiled-up water level in the core, core uncovery time, depressurization of the reactor pressure vessel, core heat-up, onset of clad oxidation, hydrogen generation, and onset of fuel relocation. The results are compared with those obtained running the MAAP 3.0B code. This PC-based computational aid can be used to train plant personnel in understanding severe accident phenomena and to assist them in managing severe accidents. (author)

  5. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  6. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  7. THALES, Thermohydraulic LOCA Analysis of BWR and PWR

    International Nuclear Information System (INIS)

    ABE, Kiyoharu

    1990-01-01

    1 - Description of program or function: THALES, which stands for 'Thermal Hydraulic Analysis of Loss-of-coolant, Emergency core cooling and Severe core damage', is a computer code system for analyzing progression of core melt accident of light water reactors. The code was developed for Level 2 PSA (probabilistic safety assessment) and applicable to a wide range of postulated accident scenarios. Its outcomes are thermal hydraulic conditions in the reactor coolant system and the containment which are necessary for analyzing fission product release and transport behavior during the accident. The code system consists of following three member codes: (1) THALES-PM for accident progression in the primary and the secondary system of PWRs, (2) THALES-BM for accident progression in the reactor coolant system of BWRs, and (3) THALES-CV for accident progression in the containment of PWRs and BWRs. The THALES-PM and the THALES-BM codes carry out two categories of analysis. The first one is overall thermal-hydraulic analysis in the reactor coolant system. The reactor coolant system is divided into multi-volumes and each volume is further separated into a liquid region and a gas region by a movable mixture level. System pressure, mixture level in each volume, coolant temperature in each region, flow rate between volumes, etc. are calculated. The other one is core heatup and meltdown analysis. The reactor core is radially and axially divided into many nodes. Fuel and cladding temperature, cladding oxidation rate, hydrogen generation rate, core melt fraction, etc. are calculated. The THALES-CV code is for containment response analysis. It divides the containment into multiple compartments, each of which is further separated into a liquid region and a gas region by a movable mixture level. Containment pressure, mixture level in each compartment, coolant temperature in each region, flow rate between compartments, etc. are calculated. The code can treat coolant blowdown from the

  8. Parametric tests of the effects of water chemistry impurities on corrosion of Zr-alloys under simulated BWR condition

    Energy Technology Data Exchange (ETDEWEB)

    Shimada, S; Ito, K [Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki (Japan); Lin, C C [GE Nucklear Energy (United States); Cheng, B [Electric Power Research Inst. (United States); Ikeda, T [Toshiba Corp. (Japan); Oguma, M [Hitachi, Ltd (Japan); Takei, T [Tokyo Electric Power Co., Inc. (Japan); Vitanza, C; Karlsen, T M [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-02-01

    The Halden BWR corrosion test loop was constructed to evaluate the impact of water chemistry variables, heat flux and boiling condition on corrosion performance of Zr-alloys in a simulated BWR environment. The loop consists of two in-core rigs, one for testing fuel rod segments and the other for evaluating water chemistry variables utilizing four miniautoclaves. Ten coupon specimens are enclosed in each miniautoclave. The Zr-alloys for the test include Zircaloy-2 having different nodular corrosion resistance and five new alloys. The first and second of the six irradiation tests planned in this program were completed. Post-irradiation examination of those test specimens have shown that the test loop is capable of producing nodular corrosion on the fuel rod cladding tested under the reference chemistry condition. The miniautoclave tests showed that nodular corrosion could be formed without flux and boiling under some water chemistry conditions and the new alloys, generally, had higher corrosion resistance than the Zircaloy in high oxygen environments. (author). 5 refs, 4 figs, 5 tabs.

  9. Cladding properties under simulated fuel pin transients

    International Nuclear Information System (INIS)

    Hunter, C.W.; Johnson, G.D.

    1975-01-01

    A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester (FCTT) to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented

  10. Analysis of corrosion behavior of KOFA cladding

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, Ki Hang; Seo, Keum Seok; Chung, Jin Gon

    1994-01-01

    The corrosion behavior of KOFA cladding was analyzed using the oxide measurement data of KOFA fuel irradiated up to the fuel rod burnup of 35,000 MWD/MTU for two cycles in Kori-2. Even though KOFA cladding is a standard Zircaloy-4 manufactured by Westinghouse according to the Siemens/KWU's HCW (Highly Cold Worked) standard Zircaloy-4 specification, it was expected that in-pile corrosion behavior of KOFA cladding would not be equivalent to that of Siemens/KWU's cladding due to the differences in such manufacturing processes as cold work and heat treatment. The analysis of measured KOFA cladding oxidation showed that oxidation of KOFA cladding is at least 19 % lower than the design analysis based upon Siemens/KWU's HCW standard Zircaloy-4 cladding. Lower corrosion of KOFA cladding seems to result from the differences in the manufacturing processes and chemical composition although the burnup and oxide layer thickness of the measured fuel rods is relatively low and the amount of the oxidation data base is small

  11. Development of advanced zirconium fuel cladding

    International Nuclear Information System (INIS)

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

    2007-04-01

    This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up (70,000MWd/mtU) which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods(LTR) in a commercial reactor

  12. Corrosion characteristics of K-claddings

    International Nuclear Information System (INIS)

    Park, J. Y.; Choi, B. K.; Jung, Y. H.; Jung, Y. H.

    2004-01-01

    The Improvement of the corrosion resistance of nuclear fuel claddings is the critical issue for the successful development of the high burn-up fuel. KAERI have developed the K-claddings having a superior corrosion resistance by controlling the alloying element addition and optimizing the manufacturing process. The comparative evaluation of the corrosion resistance for K-claddings and the foreign claddings was performed and the effect of the heat treatment on the corrosion behavior of K-claddings was also examined. Corrosion tests were carried out in the conditions of 360 .deg. C pure water, PWR-simulating loop and 400 .deg. C steam, From the results of the corrosion tests, it was found that the corrosion resistance of K-claddings is superior to those of Zry4 and A claddings and K6 showed a better corrosion resistance than K3. The corrosion behavior of K-cladding was strongly influenced by the final annealing rather than the intermediate annealing, and the corrosion resistance increased with decreasing the final annealing temperature

  13. Cladding Alloys for Fluoride Salt Compatibility

    Energy Technology Data Exchange (ETDEWEB)

    Muralidharan, Govindarajan [ORNL; Wilson, Dane F [ORNL; Walker, Larry R [ORNL; Santella, Michael L [ORNL; Holcomb, David Eugene [ORNL

    2011-06-01

    This report provides an overview of several candidate technologies for cladding nickel-based corrosion protection layers onto high-temperature structural alloys. The report also provides a brief overview of the welding and weld performance issues associated with joining nickel-clad nickel-based alloys. From the available techniques, two cladding technologies were selected for initial evaluation. The first technique is a line-of-sight method that would be useful for cladding large structures such as vessel interiors or large piping. The line-of-sight method is a laser-based surface cladding technique in which a high-purity nickel powder mixed into a polymer binder is first sprayed onto the surface, baked, and then rapidly melted using a high-power laser. The second technique is a vapor phase technique based on the nickel-carbonyl process that is suitable for cladding inaccessible surfaces such as the interior surfaces of heat exchangers. An initial evaluation for performed on the quality of nickel claddings processed using the two selected cladding techniques.

  14. Best-estimate analysis development for BWR systems

    International Nuclear Information System (INIS)

    Sutherland, W.A.; Alamgir, M.; Kalra, S.P.; Beckner, W.D.

    1986-01-01

    The Full Integral Simulation Test (FIST) Program is a three pronged approach to the development of best-estimate analysis capability for BWR systems. An experimental program in the FIST BWR system simulator facility extends the LOCA data base and adds operational transients data. An analytical method development program with the BWR-TRAC computer program extends the modeling of BWR specific components and major interfacing systems, and improves numerical techniques to reduce computer running time. A method qualification program tests TRAC-B against experiments run in the FIST facility and extends the results to reactor system applications. With the completion and integration of these three activities, the objective of a best-estimate analysis capability has been achieved. (author)

  15. Boiling water system of nuclear power plants (BWR)

    International Nuclear Information System (INIS)

    Martias Nurdin

    1975-01-01

    About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

  16. BWR stability: history and state-of-the-art

    International Nuclear Information System (INIS)

    Yadigaroglu, George

    2014-01-01

    The paper briefly recalls the historical developments, reviews the important phenomena, the analytical and simulation tools that are used for the analysis of BWR stability focussing on the linear, frequency domain methods

  17. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  18. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

  19. Development of long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Maruya, T.; Hiraiwa, K.; Arai, K.; Narabayash, T.; Aritomi, M.

    2002-01-01

    This paper describes an innovative plant concept for long operating cycle simplified BWR (LSBWR) In this plant concept, 1) Long operating cycle ( 3 to 15 years), 2) Simplified systems and building, 3) Factory fabrication in module are discussed. Designing long operating core is based on medium enriched U-235 with burnable poison. Simplified systems and building are realized by using natural circulation with bottom located core, internal CRD and PCV with passive system and an integrated reactor and turbine building. This LSBWR concept will have make high degree of safety by IVR (In Vessel Retention) capability, large water inventory above the core region and no PCV vent to the environment due to PCCS (Passive Containment Cooling System) and internal vent tank. Integrated building concept could realize highly modular arrangement in hull structure (ship frame structure), ease of seismic isolation capability and high applicability of standardization and factory fabrication. (authors)

  20. BWR plant advanced central control panel PODIA

    International Nuclear Information System (INIS)

    Fujii, K.; Hayakawa, H.; Ikeda, Y.; Neda, T.; Suto, O.; Takamiya, S.

    1983-01-01

    BWR plant central control panels have become more and more enlarged and complicated recently due to the magnification of the scale of a plant and the requirement to reinforce safety. So, it is important to make communication between men and the complicated central control panel smooth. Toshiba has developed an advanced central control panel, named PODIA, which uses many computers and color CRTs, and PODIA is now in the stage of application to practical plants. In this article, the writers first touch upon control functions transition in the central control room, the PODIA position concerning the world-wide trend in this technology phase and the human engineering on the design. Then they present concrete design concepts for the control board and computer system which constitute PODIA

  1. Evaluation of internal flooding in a BWR

    International Nuclear Information System (INIS)

    Shiu, K.; Papazoglou, I.A.; Sun, Y.H.; Anavim, E.; Ilberg, D.

    1985-01-01

    Flooding inside a nuclear power station is capable of concurrently disabling redundant safety systems. This paper presents the results of a recent review study performed on internally-generated floods inside a boiling water reactor (BWR) reactor building. The study evaluated the flood initiator frequency due to either maintenance or ruptures using Markovian models. A time phased event tree approach was adopted to quantify the core damage frequency based on the flood initiator frequency. It is found in the study that the contribution to the total core damage due to internal flooding events is not insignificant and is comparable to other transient contributors. The findings also indicate that the operator plays an important role in the prevention as well as the mitigation of a flooding event

  2. Seismic risk assessment of a BWR

    International Nuclear Information System (INIS)

    Wells, J.E.; Bernreuter, D.L.; Chen, J.C.; Lappa, D.A.; Chuang, T.Y.; Murray, R.C.; Johnson, J.J.

    1987-01-01

    The simplified seismic risk methodology developed in the USNRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant (PWR). The simplified seismic risk methodology was developed to reduce the costs associated with a seismic risk analysis while providing adequate results. A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models, was developed and used in assessing the seismic risk of the Zion nuclear power plant (FSAR). The simplified seismic risk methodology was applied to the LaSalle County Station nuclear power plant, a BWR; to further demonstrate its applicability, and if possible, to provide a basis for comparing the seismic risk from PWRs and BWRs. (orig./HP)

  3. A BWR Safety and Operability Improvements

    International Nuclear Information System (INIS)

    Sawyer, Craig D.

    1993-01-01

    The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

  4. BWR stability using a reducing dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J. M.; Blazquez Martinez, J. B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  5. BWR stability using a reduced dynamical model

    International Nuclear Information System (INIS)

    Ballestrin Bolea, J.M.; Blazquez, J.B.

    1990-01-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical struct-ure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations in non-linear. Simple parametric calculat-ion of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author). 7 refs

  6. Advanced technology for BWR operator training simulator

    International Nuclear Information System (INIS)

    Shibuya, Akira; Fujita, Eimitsu; Nakao, Toshihiko; Nakabaru, Mitsugu; Asaoka, Kouchi.

    1991-01-01

    This paper describes an operator training simulator for BWR nuclear power plants which went into service recently. The simulator is a full scope replica type simulator which faithfully replicates the control room environment of the reference plant with six main control panels and twelve auxiliary ones. In comparison with earlier simulators, the scope of the simulation is significantly extended in both width and depth. The simulation model is also refined in order to include operator training according to sympton-based emergency procedure guidelines to mitigate the results in accident cases. In particular, the core model and the calculational model of the radiation intensity distribution, if radioactive materials were released, are improved. As for simulator control capabilities by which efficient and effective training can be achieved, various advanced designs are adopted allowing easy use of the simulators. (author)

  7. Manufacturing technology and process for BWR fuel

    International Nuclear Information System (INIS)

    Kato, Shigeru

    1996-01-01

    Following recent advanced technologies, processes and requests of the design changes of BWR fuel, Nuclear Fuel Industries, Ltd. (NFI) has upgraded the manufacturing technology and honed its own skills to complete its brand-new automated facility in Tokai in the latter half of 1980's. The plant uses various forms of automation throughout the manufacturing process: the acceptance of uranium dioxide powder, pelletizing, fuel rod assembling, fuel bundle assembling and shipment. All processes are well computerized and linked together to establish the integrated control system with three levels of Production and Quality Control, Process Control and Process Automation. This multi-level system plays an important role in the quality assurance system which generates the highest quality of fuels and other benefits. (author)

  8. Facility of BWR type nuclear power plant

    International Nuclear Information System (INIS)

    Kubo, Mitsuji

    1998-01-01

    A condensate filtering device for cleaning condensate flown from a low pressure turbine and a condensate desalting device are connected by way of a condensate pipeline. Control rod drives (CRD) are disposed to the lower portion of BWR. A CRD pump and one end of a CRD feedwater pipeline are connected in series to the upstream of CRD. The other end of the CRD feedwater pipeline is connected to a CRD water taking pipeline branched from the condensate pipeline. Water is taken to the CRD from downstream of the condensate filtering device and upstream of a connecting portion between a low pressure heater drain pipeline and the condensate pipeline. Flow of impurities leached out of the condensate desalting device to the reactor can be suppressed, and rising of temperature of CRD water by the low pressure heater drain water is prevented. In addition, flowing of dissolved oxygen to the CRD system can be suppressed. (I.N.)

  9. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Petek, M.

    1991-01-01

    A recently completed Oak Ridge effort proposes two management strategies for mitigation of the events that might occur in-vessel after the onset of significant core damage in a BWR severe accident. While the probability of such an accident is low, there may be effective yet inexpensive mitigation measures that could be implemented employing the existing plant equipment and requiring only additions to the plant emergency procedures. In this spirit, accident management strategies have been proposed for use of a borated solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and for containment flooding to maintain the core debris within the reactor vessel if injection systems cannot be restored. The proposed strategy for poisoning of the water used for vessel reflood should injection systems be restored after control blade damage has occurred has great promise, using only the existing plant equipment but employing a different chemical form for the boron poison. The dominant BWR severe accident sequence is Station Blackout and without means for mechanical stirring or heating of the storage tank, the question of being able to form the poisoned solution under accident conditions becomes of supreme importance. On the other hand, the proposed strategy for drywell flooding to cool the reactor vessel bottom head and prevent the core and structure debris from escaping to the drywell holds less promise. This strategy does, however, have potential for future plant designs in which passive methods might be employed to completely submerge the reactor vessel under severe accident conditions without the need for containment venting

  10. Seismic PRA of a BWR plant

    International Nuclear Information System (INIS)

    Nishio, Masahide; Fujimoto, Haruo

    2014-01-01

    Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

  11. BWR startup and shutdown activity transport control

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

    2010-07-01

    This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

  12. Flaw behavior in mechanically loaded clad plates

    International Nuclear Information System (INIS)

    Iskander, S.K.; Robinson, G.C.; Oland, C.B.

    1989-01-01

    A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

  13. Some proposed mechanisms for internal cladding corrosion

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Pickering, S.; Whitlow, W.H.

    1977-01-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  14. Some proposed mechanisms for internal cladding corrosion

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M H; Pickering, S; Whitlow, W H [EURATOM (United Kingdom)

    1977-04-01

    In spite of extensive research during recent years, a comprehensive model for internal cladding corrosion in fast reactor oxide fuel pins has not yet been established. In this paper, a model is proposed which accounts for many of the features normally associated with this type of corrosion. The model is composed of a number of parts which describe the chronological sequence of events at the fuel/cladding interface. The corrosion reaction is visualised as being primarily chemical in character, involving the cladding steel, the fuel and the more aggressive fission products, notably caesium in the presence of oxygen. The model attempts to explain how corrosion starts, how it depends on the oxygen potential, why it occurs non-uniformly; also covered are phase changes within the cladding steel and morphological features such as the intergranular form of attack and the distribution of corrosion products in the fuel/cladding gap. (author)

  15. Modernity after Modernity

    Directory of Open Access Journals (Sweden)

    Marin Dinu

    2007-08-01

    Full Text Available A strategy for the second modernization raises, beyond objectives, a series of epistemicresponsibilities. It is known that modernization stemming from the Enlightment had, among other things,the pretense that it is a project which is self-legitimating. Its profound rationales are the only justification.Referential self-centering proved to be the one that made possible a practice of the new. Modernizationhaving the function of renouncing myth – meaning an eliminatory formula for the past – and thefixation in the opportunity and potentiality of the present, seemed to close an insoluble but extremelyengrossing problem: that of a propensity towards utopia, of the risky escape towards the future. Thetraditionalization of the new constitutes a support for the daring to break out of the captivity of themoment.Modernization becomes the experience of combining the new which, thus, creates a succession ofpresent times. The future is no longer the result of fantasy, but a system’s direct expression to combine thenew. Therefore the future is an option for one or another model of the present, often tested previouslysomewhere else. In a non-metaphysical way, the future can be seen, touched, tried, lived by simplegeographical movement. The sense of evolution has de-temporalized taking the form of the concomitant,parallel, enclosed, neighboring space. We just have to be in the trend, to evolve in the context.Globalization defines the context and its conception – as a project of the second modernity – showsus the trends. The problem is how to understand the context in order to find the sense of the trend. Are wethe load the sense with the values of the first modernity or will we have to turn to the values of anothermodernity? Why do we have to move away from the significance of the processes which made up the firstmodernity? How do we relate to the content of the new context in which the structural trends of today’sworld are taking place? What is the

  16. BWR Radiation Assessment and Control Program: assessment and control of BWR radiation fields. Volume 1. Executive summary

    International Nuclear Information System (INIS)

    Anstine, L.D.

    1983-05-01

    This report covers work on the BWR Radiation Assessment and Control (BRAC) Program from 1978 to 1982. The major activities during this report period were assessment of the radiation-level trends in BWRs, evaluation of the effects of forward-pumped heater drains on BWR water quality, installation and operation of a corrosion-product deposition loop in an operating BWR, and analyzation of fuel-deposit samples from two BWRs. Radiation fields were found to be controlled by cobalt-60 and to vary from as low as 50 mr/hr to as high as 800 mr/hr on the recirculation-system piping. Detailed information on BWR corrosion films and system deposits is presented in the report. Additionally, the results of an oxygen-injection experiment and recontamination monitoring studies are provided

  17. Behaviour of the reactivity for BWR fuel cells; Comportamiento de la reactividad para celdas de combustible BWR

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, J. A.; Alonso, G.; Delfin, A.; Vargas, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: galonso@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, U. P. Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico)

    2011-11-15

    In this work the behaviour of the reactivity of a fuel assembly type BWR was studied, the objective is to obtain some expressions that consider the average enrichment of U-235 and the gadolinium concentration like a function of the fuel cells burnt. Also, the applicability of the lineal reactivity model was analyzed for fuel cells type BWR. The analysis was carried out with the CASMO-4 code. (Author)

  18. Modern teaching for modern education

    OpenAIRE

    Mirascieva, Snezana

    2016-01-01

    Carrying the epithet of being contemporary education today means modern teaching. If modern education is a state in the field of education of all its elements, then teaching will also be a state with its own special features defining it as modern. The main issues of concern in this paper relate to what constitutes modern teaching, which features determine it as being modern, and how much is teaching today following the trend of modernization.

  19. Metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, Nina

    This work concerns planar optical waveguide sensors for biosensing applications, with the focus on deep-probe sensing for micron-scale biological objects like bacteria and whole cells. In the last two decades planar metal-clad waveguides have been brieflyintroduced in the literature applied...... for various biosensing applications, however a thorough study of the sensor configurations has not been presented, but is the main subject of this thesis. Optical sensors are generally well suited for bio-sensing asthey show high sensitivity and give an immediate response for minute changes in the refractive...... index of a sample, due to the high sensitivity of optical bio-sensors detection of non-labeled biological objects can be performed. The majority of opticalsensors presented in the literature and commercially available optical sensors are based on evanescent wave sensing, however most of these sensors...

  20. Advanced Construction of Compact Containment BWR

    International Nuclear Information System (INIS)

    Takahashi, M.; Maruyama, T.; Mori, H.; Hoshino, K.; Hijioka, Y.; Heki, H.; Nakamaru, M.; Hoshi, T.

    2006-01-01

    The reactor concept considered in this paper has a mid/small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. Compact Containment BWR (CCR) is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified safety system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response time in case of design base accidents including loss of coolant accidents. The IC suppresses PCV pressure by steam condensation without any AC power. Cooling the molten core inside the RPV if the core should be damaged by loss of core coolability could attain the IVR. CCR's specific self-standing steel high pressure resistible PCV is designed to contain minimum piping and valves inside with reactor pressure vessel (RPV), only 13 m in diameter and 24 m in height. This compact PCV makes it possible to

  1. Fuel cladding mechanical interaction during power ramps

    International Nuclear Information System (INIS)

    Guerin, Y.

    1985-01-01

    Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i) fuel swelling especially at levels of caesium accumulation, and ii) thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction (FCMI). Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients

  2. Strain-induced corrosion cracking in ferritic components of BWR primary circuits

    International Nuclear Information System (INIS)

    Seifert, H.-P.; Ritter, S.; Ineichen, U.; Tschanz, U.; Gerodetti, B.

    2003-04-01

    The present final report of the RIKORR project is a summary of a literature survey and of the experimental work performed by PSI on the environmentally-assisted cracking (EAC) and dynamic strain ageing (DSA) susceptibility of low-alloy steels (LAS) in high-temperature (HT) water. Within this project, the EAC crack growth behaviour of different low-alloy RPV steels, weld filler and weld heat-affected zone materials has been investigated under simulated transient and steady-state BWR/NWC power operation conditions. The strain-induced corrosion cracking (SICC) / low-frequency corrosion fatigue (CF) and stress corrosion cracking (SCC) crack growth behaviour of different low-alloy RPV steels under simulated transient and stationary BWR/NWC conditions was characterized by slow rising load / low-frequency corrosion fatigue and constant load / periodical partial unloading / ripple load tests with pre-cracked fracture mechanics specimens in oxygenated HT water at temperatures of either 288, 250, 200 or 150 o C. Modern high-temperature water loops, on-line crack growth monitoring and fractographic analysis by scanning electron microscopy (SEM) were used to quantify the cracking response. (author)

  3. Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

    Energy Technology Data Exchange (ETDEWEB)

    Ernst, Frank [Case Western Reserve Univ., Cleveland, OH (United States)

    2016-10-13

    We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute – carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress–corrosion cracking (hydrogen-induced embrittlement), and – potentially – radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non–treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

  4. Improved Accident Tolerance of Austenitic Stainless Steel Cladding through Colossal Supersaturation with Interstitial Solutes

    International Nuclear Information System (INIS)

    Ernst, Frank

    2016-01-01

    We proposed a program-supporting research project in the area of fuel-cycle R&D, specifically on the topic of advanced fuels. Our goal was to investigate whether SECIS (surface engineering by concentrated interstitial solute - carbon, nitrogen) can improve the properties of austenitic stainless steels and related structural alloys such that they can be used for nuclear fuel cladding in LWRs (light-water reactors) and significantly excel currently used alloys with regard to performance, safety, service life, and accident tolerance. We intended to demonstrate that SECIS can be adapted for post-processing of clad tubing to significantly enhance mechanical properties (hardness, wear resistance, and fatigue life), corrosion resistance, resistance to stress-corrosion cracking (hydrogen-induced embrittlement), and - potentially - radiation resistance (against electron-, neutron-, or ion-radiation damage). To test this hypothesis, we measured various relevant properties of the surface-engineered alloys and compared them with corresponding properties of the non-treated, as-received alloys. In particular, we studied the impact of heat exposure corresponding to BWR (boiling-water reactor) working and accident (loss-of-coolant) conditions and the effect of ion irradiation.

  5. For the world's best cladding tubes, ten years of progress by Zircaloy Special Committee of JAPCO

    International Nuclear Information System (INIS)

    Mishima, Yoshitsugu

    1982-01-01

    The zircaloy special committee was organized in 1971 for the purpose of planning the trial use of two nuclear fuel assemblies for which Japan-made cladding tubes were to be used, for a BWR. Now, seven years later, these two fuel assemblies have completed their service life, and have been submitted to post-irradiation examination after cooling for a year. Zircaloy tubes have been produced by Sumitomo Metal Industries, Ltd., and Kobe Steel, Ltd., and more than ten years have elapsed since wholly Japan-made zircaloy cladding tubes were used for reloading fuel elements for the Japan Power Demonstration Reactor. In this report, the history, progress and significance of the works performed by the committee are summarized. The LWR fuel elements made in Japan have attained the highest performance in the world as the leak has been scarce, and the works of the committee is one of the pioneering activities in the development of LWR fuel technology. The situation for starting the committee, the activity of the committee during ten years, the significance and outcome of the committee activity are reported. (Kako, I.)

  6. OECD/NRC BWR Turbine Trip Benchmark: Simulation by POLCA-T Code

    International Nuclear Information System (INIS)

    Panayotov, Dobromir

    2004-01-01

    Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and three-dimensional (3-D) neutron kinetics core models. Participation in the OECD/NRC BWR Turbine Trip (TT) Benchmark is a part of our efforts toward the code's validation. The paper describes the objectives for TT analyses and gives a brief overview of the developed plant system input deck and 3-D core model.The results of exercise 1, system model without netronics, are presented. Sensitivity studies performed cover the maximal time step, turbine stop valve position and mass flow, feedwater temperature, and steam bypass mass flow. Results of exercise 2, 3-D core neutronic and thermal-hydraulic model with boundary conditions, are also presented. Sensitivity studies include the core inlet temperature, cladding properties, and direct heating to core coolant and bypass.The entire plant model was validated in the framework of the benchmark's phase 3. Sensitivity studies include the effect of SCRAM initialization and carry-under. The results obtained - transient fission power and its initial axial distribution and steam dome, core exit, lower and upper plenum, main steam line, and turbine inlet pressures - showed good agreement with measured data. Thus, the POLCA-T code capabilities for correct simulation of pressurizing transients with very fast power were proved

  7. Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Area de Ingenieria en Recursos Energeticos, Iztapalapa (Mexico)

    2014-07-01

    A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

  8. Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

    International Nuclear Information System (INIS)

    Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

    2014-01-01

    A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

  9. Development status of compact containment BWR

    International Nuclear Information System (INIS)

    Heki, H.; Nakamaru, M.; Mori, H.; Sekiguchi, K.; Kuroki, M.; Arai, K.; Hida, T.

    2005-01-01

    In Japan, increase of nuclear plant unit capacity has been promoted to take advantage of economies of scale while further enhancing safety and reliability. As a result, more than 50 units of nuclear power plants are playing important role in electric power generation. However, the factors, such as stagnant growth in the recent electricity demand, limitation in electricity grid capacity and limited in initial investment avoiding risk, will not be in favor of large plant outputs. The reactor concept considered in this paper has a small power output, a compact containment and a simplified BWR configuration with comprehensive safety features. The Compact Containment Boiling Water Reactor (CCR), which is being developed with matured BWR technologies together with innovative systems/components, will provide attractiveness for the energy market in the world due to its flexibility in energy demands as well as in site conditions, its high potential in reducing investment risk and its safety feature facilitating public acceptance. The flexibility is achieved by CCR's mid/small power output of 400 MWe class and capability of long operating cycle (refueling intervals). The high investment potential is expected from CCR's simplification/innovation in design such as natural circulation core cooling with the bottom located short core, top mounted upper entry control rod drives (CRDs) with ring-type dryers and simplified ECCS system with high pressure resistible primary containment vessel (PCV) concept. The natural circulation core eliminates recirculation pumps as well as needs for maintenance of such pumps. The top mounted upper entry CRDs enable the bottom located short core in RPV. The safety feature mainly consists of large water inventory above the core without large penetration below the top of the core, passive cooling system by isolation condenser (IC), high pressure resistible PCV and in-vessel retention (IVR) capability. The large inventory increases the system response

  10. Comparison of the CORA-12, 13, 17 experiments and B4 effect on the flooding behavior of BWR bundles

    International Nuclear Information System (INIS)

    Hagen, S.; Sepold, L.; Wallenfels, K.P.; Hofmann, P.; Noack, V.; Schanz, G.; Schumacher, G.

    1995-01-01

    The CORA quench experiments 12, 13 (PWR) and 17 (BWR) are in agreement with LOFT 2 and TMI: Flooding of hot Zircaloy clad fuel rods does not result in an immediate cooldown of the bundle, but produces remarkable temporary temperature increase, connected to a strong peak in hydrogen production. The PWR tests CORA 12 and CORA 13 are of the same geometrical arrangement and test conduct, with the exception of the shorter time between power shutdown and quench initiation for CORA 13. A higher temperature of the bundle at start of quenching was the consequence. BWR test CORA 17 - with B 4 C absorber and additional Zircaloy channel box walls - was in respect to the delay-time between power shutdown and start of quenching similar to test CORA 12. All tests showed during the quench phase the temporary temperature increase, correlated to a hydrogen peak. The CORA 17 test resulted immediately after quenching in a modest increase for 20 s and changed then in a steep increase, resulting in the highest temperature and hydrogen peaks of the three tests. CORA 17 also showed a temperature increase in the lower part of the bundle, in contrast to CORA 12 and CORA 13 with temperature increase only in the upper half of the bundle. We interpret this earlier starting and stronger reaction due to the influence of the boron carbide, the absorber material of the BWR test. B 4 C has an exothermic reaction rate 4 to 9 times larger than Zry and produces 5 to 6,6 times more hydrogen. Probably the hot remained columns of B 4 C (seen in the non-quench test CORA 16) react early in the quench process with the increased upcoming steam. The bundle temperature raised by this reaction increases the reaction rate (exponential dependency) of the remaining metallic Zry. Due to the larger amount of Zry in the BWR bundle (channel box walls) and the smaller steam input during the heatup phase (2 g/s instead of 6 g/s) more metallic Zry can have survived oxidation during the heatup phase. (orig./HP)

  11. Chemical compatibility between cladding alloys and advanced fuels

    International Nuclear Information System (INIS)

    Fee, D.C.; Johnson, C.E.

    1975-05-01

    The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium (versus gas) bonded fuels. The depth of carburization increases with increasing sesquicarbide (M 2 C 3 ) content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. (U.S.)

  12. Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Larry J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate

  13. Friction Surface Cladding of AA1050 on AA2024-T351; influence of clad layer thickness and tool rotation rate

    NARCIS (Netherlands)

    Liu, Shaojie; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko

    2015-01-01

    Friction Surfacing Cladding (FSC) is a recently developed solid state process to deposit thin metallic clad layers on a substrate. The process employs a rotating tool with a central opening to supply clad material and support the distribution and bonding of the clad material to the substrate. The

  14. Water level measurement uncertainty during BWR instability

    International Nuclear Information System (INIS)

    Torok, R.C.; Derbidge, T.C.; Healzer, J.M.

    1994-01-01

    This paper addresses the performance of the water-level measurement system in a boiling water reactor (BWR) during severe instability oscillations which, under some circumstances, can occur during an anticipated transient without SCRAM (ATWS). Test data from a prototypical mock-up of the water-level measurement system was used to refine and calibrate a water-level measurement system model. The model was then used to predict level measurement system response, using as boundary conditions vessel pressures calculated by ppercase RETRAN for an ATWS/instability event.The results of the study indicate that rapid pressure changes in the reactor pressure vessel which cause oscillations in downcomer water level, coupled with differences in instrument line lengths, can produce errors in the sensed water level. Using nominal parameters for the measurement system components, a severe instability transient which produced a 0.2 m peak-to-minimum water-level oscillation in the vessel downcomer was predicted to produce pressure difference equivalent to a 0.7 m level oscillation at the input to the differential pressure transmitter, 0.5 m oscillation at the output of the transmitter, and an oscillation of 0.3 m on the water-level indicator in the control room. The level measurement system error, caused by downcomer water-level oscillations and instrument line length differential, is mitigated by damping both in the differential pressure transmitter used to infer level and in the control room display instrument. ((orig.))

  15. Recent technology for BWR operator training simulators

    International Nuclear Information System (INIS)

    Sato, Takao; Hashimoto, Shigeo; Kato, Kanji; Mizuno, Toshiyuki; Asaoka, Koichi.

    1990-01-01

    As one of the important factors for maintaining the high capacity ratio in Japanese nuclear power stations, the contribution of excellent operators is pointed out. BWR Operation Training Center has trained many operators using two full scope simulators for operation training modeling BWRs. But in order to meet the demands of the recent increase of training needs and the upgrading of the contents, it was decided to install the third simulator, and Hitachi Ltd. received the order to construct the main part, and delivered it. This simulator obtained the good reputation as its range of simulation is wide, and the characteristics resemble very well those of the actual plants. Besides, various new designs were adopted in the control of the simulator, and its handling became very easy. Japanese nuclear power plants are operated at constant power output, and the unexpected stop is very rare, therefore the chance of operating the plants by operators is very few. Accordingly, the training using the simulators which can simulate the behavior of the plants with computers, and can freely generate abnormal phenomena has become increasingly important. The mode and positioning of the simulators for operation training, the full scope simulator BTC-3 and so on are reported. (K.I.)

  16. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  17. Method of operating BWR type power plants

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To improve the operation efficiency of BWR type reactors by reducing the time from the start-up of the reactor to the start-up of the turbine and electrical generator, as well as decrease the pressure difference in each of the sections of the pressure vessel to thereby extend its life span. Method: The operation comprises switching the nuclear reactor from the shutdown mode to the start-up mode, increasing the reactor power to a predetermined level lower than a rated power while maintaining the reactor pressure to a predetermined level lower than a rated pressure, starting up a turbine and an electrical generator in the state of the predetermined reactor pressure and the reactor power to connect the electrical generator to the power transmission system and, thereafter, increasing the reactor pressure and the reactor power to the predetermined rated pressure and rated power respectively. This can shorten the time from the start-up of the reactor to the start of the power transmission system, whereby the operation efficiency of the power plant can be improved. (Moriyama, K.)

  18. Experience on a BWR plant diagnosis system

    International Nuclear Information System (INIS)

    Tanabe, A.; Kawai, K.; Hashimoto, Y.

    1981-01-01

    It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

  19. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  20. Fuel assemblies for BWR type reactors

    International Nuclear Information System (INIS)

    Ishizuka, Takao.

    1981-01-01

    Purpose: To enable effective failed fuel detection by the provision of water rod formed with a connecting section connected to a warmed water feed pipe of a sipping device at the lower portion and with a warmed water jetting port in the lower portion in a fuel assembly of a BWR type reactor to thereby carry out rapid sipping. Constitution: Fuel rods and water rods are contained in the channel box of a fuel assembly, and the water rod is provided at its upper portion with a connecting section connected to the warmed water feed pipe of the sipping device and formed at its lower portion with a warmed water jetting port for jetting warmed water fed from the warmed water feed pipe. Upon detection of failed fuels, the reactor operation is shut down and the reactor core is immersed in water. The cover for the reactor container is removed and the cap of the sipping device is inserted to connect the warmed water feed pipe to the connecting section of the water rod. Then, warmed water is fed to the water rod and jetted out from the warmed water jetting port to cause convection and unify the water of the channel box in a short time. Thereafter, specimen is sampled and analyzed for the detection of failed fuels. (Moriyama, K.)

  1. Primary coolant system of BWR type reactor

    International Nuclear Information System (INIS)

    Ibe, Hidefumi; Takahashi, Masanori; Aoki, Yasuko

    1997-01-01

    The present invention provides a water quality control system for preventing corrosion and for extending working life of structural materials of a BWR-type reactor. Namely, a sensor group 1 and a sensor group 2 are disposed at different positions such as in a feedwater system, a recycling system, main steam pipes, and a pressure vessel, respectively. Each sensor group can record and generate alarms independently. The sensor group 1 for usual monitoring is connected to a calculation device by way of a switch to confirm that the monitored values are within a proper range by the injection of a water quality moderating agent. The sensor group 2 is caused to stand alone or connected with the calculation device by way of a switch optionally. When abnormality should occur in the sensor group 1, the sensor group 2 determines the limit for the increase/decrease of controlling amount of the moderating agent at a portion where the conditions are changed to the most severe direction by using data base. The moderating agent is injected and controlled based on the controlling amount. The system of the present invention can optionally cope with a new sensor and determination for new water quality standards. Then the evaluation/control accuracy of the entire system can be improved while covering up the errors of each sensor. (I.S.)

  2. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  3. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  4. MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION

    Directory of Open Access Journals (Sweden)

    Miroslav Cech

    2016-12-01

    Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.

  5. Method of processing spent fuel cladding tubes

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

    1986-01-01

    Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

  6. GSGG edge cladding development: Final technical report

    International Nuclear Information System (INIS)

    Izumitani, T.; Meissner, H.E.; Toratani, H.

    1986-01-01

    The objectives of this project have been: (1) Investigate the possibility of chemical etching of GSGG crystal slabs to obtain increased strength. (2) Design and construct a simplified mold assembly for casting cladding glass to the edges of crystal slabs of different dimensions. (3) Conduct casting experiments to evaluate the redesigned mold assembly and to determine stresses as function of thermal expansion coefficient of cladding glass. (4) Clad larger sizes of GGG slabs as they become available. These tasks have been achieved. Chemical etching of GSGG slabs does not appear possible with any other acid than H 3 PO 4 at temperatures above 300 0 C. A mold assembly has been constructed which allowed casting cladding glass around the edges of the largest GGG slabs available (10 x 20 x 160 mm) without causing breakage through the annealing step

  7. Duplex stainless steel surface bay laser cladding

    International Nuclear Information System (INIS)

    Amigo, V.; Pineda, Y.; Segovia, F.; Vicente, A.

    2004-01-01

    Laser cladding is one of the most promising techniques to restore damaged surfaces and achieve properties similar to those of the base metal. In this work, duplex stainless steels have been cladded by a nickel alloy under different processing conditions. The influence of the beam speed and defocusing variables ha been evaluated in the microstructure both of the cladding and heat affected zone, HAZ. These results have been correlated to mechanical properties by means of microhardness measurements from cladding area to base metal through the interface. This technique has shown to be very appropriate to obtain controlled mechanical properties as they are determined by the solidification microstructure, originated by the transfer of mass and heat in the system. (Author) 21 refs

  8. Corrosion behaviour of cladded nickel base alloys

    International Nuclear Information System (INIS)

    Brandl, W.; Ruczinski, D.; Nolde, M.; Blum, J.

    1995-01-01

    As a consequence of the high cost of nickel base alloys their use as surface layers is convenient. In this paper the properties of SA-as well as RES-cladded NiMo 16Cr16Ti and NiCr21Mo14W being produced in single and multi-layer technique are compared and discussed with respect to their corrosion behaviour. Decisive criteria describing the qualities of the claddings are the mass loss, the susceptibility against intergranular corrosion and the pitting corrosion resistance. The results prove that RES cladding is the most suitable technique to produce corrosion resistant nickel base coatings. The corrosion behaviour of a two-layer RES deposition shows a better resistance against pitting than a three layer SAW cladding. 7 refs

  9. CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

    International Nuclear Information System (INIS)

    P. Macheret

    2001-01-01

    In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points pertaining to irradiated Zircaloy cladding. The scope of the calculation is provided by the range of experimental parameters characterized by Zircaloy cladding temperature between 292 C and 420 C, hoop stress between 50 and 630 MPa, and test time extending to 8000 h. As for the burnup of the experimental samples, it ranges between 0.478 and 64 MWd/kgU (i.e., megawatt day per kilogram of uranium), but this is not a parameter of the adapted correlation

  10. Burnup credit feasibility for BWR spent fuel shipments

    International Nuclear Information System (INIS)

    Broadhead, B.L.

    1990-01-01

    Considerable interest in the allowance of reactivity credit for the exposure history of power reactor fuel currently exists. This ''burnup credit'' issue has the potential to greatly reduce risk and cost when applied to the design and certification of spent of fuel casks used for transportation and storage. Analyses 1 have shown the feasibility estimated the risk and economic incentives for allowing burnup credit in pressurized water reactor (PWR) spent fuel shipping cask applications. This paper summarizes the extension of the previous PWR feasibility assessments to boiling water reactor (BWR) fuel. As with the PWR analysis, the purpose was not verification of burnup credit (see ref. 2 for ongoing work in this area) but a reasonable assessment of the feasibility and potential gains from its use in BWR applications. This feasibility analysis aims to apply simple methods that adequately characterize the time-dependent isotopic compositions of typical BWR fuel. An initial analysis objective was to identify a simple and reliable method for characterizing BWR spent fuel. The method includes characterization of a typical pin-cell spectrum, using a one-dimensional (1-D) model of a BWR assembly. The calculated spectrum allows burnup-dependent few-group material constants to be generated. Point depletion methods were then used to obtain the time-varying characteristics of the fuel. These simple methods were validated, where practical, with multidimensional methods. 6 refs., 1 tab

  11. Optimization of metal-clad waveguide sensors

    DEFF Research Database (Denmark)

    Skivesen, N.; Horvath, R.; Pedersen, H.C.

    2005-01-01

    The present paper deals with the optimization of metal-clad waveguides for sensor applications to achieve high sensitivity for adlayer and refractive index measurements. By using the Fresnel reflection coefficients both the angular shift and the width of the resonances in the sensorgrams are taken...... into account. Our optimization shows that it is possible for metal-clad waveguides to achieve a sensitivity improvement of 600% compared to surface-plasmon-resonance sensors....

  12. BWR Spent Nuclear Fuel Interfacial Bonding Efficiency Study

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jiang, Hao [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-04-30

    The objective of this project is to perform a systematic study of spent nuclear fuel (SNF, also known as “used nuclear fuel” [UNF]) integrity under simulated transportation environments using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) hot-cell testing technology developed at Oak Ridge National Laboratory (ORNL) in August 2013. Under Nuclear Regulatory Commission (NRC) sponsorship, ORNL completed four benchmark tests, four static tests, and twelve dynamic or cycle tests on H. B. Robinson (HBR) high burn-up (HBU) fuel. The clad of the HBR fuels was made of Zircaloy-4. Testing was continued in fiscal year (FY) 2014 using Department of Energy (DOE) funds. Additional CIRFT testing was conducted on three HBR rods; two specimens failed, and one specimen was tested to over 2.23 × 107 cycles without failing. The data analysis on all the HBR SNF rods demonstrated that it is necessary to characterize the fatigue life of the SNF rods in terms of (1) the curvature amplitude and (2) the maximum absolute of curvature extremes. The maximum extremes are significant because they signify the maximum tensile stress for the outer fiber of the bending rod. CIRFT testing has also addressed a large variation in hydrogen content on the HBR rods. While the load amplitude is the dominant factor that controls the fatigue life of bending rods, the hydrogen content also has an important effect on the lifetime attained at each load range tested. In FY 15, eleven SNF rod segments from the Limerick BWR were tested using the ORNL CIRFT equipment; one test under static conditions and ten tests under dynamic loading conditions. Under static unidirectional loading, a moment of 85 N·m was obtained at a maximum curvature of 4.0 m-1. The specimen did not show any sign of failure during three repeated loading cycles to a similar maximum curvature. Ten cyclic tests were conducted with amplitudes varying from 15.2 to 7.1 N·m. Failure was observed in nine of

  13. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  14. Application of gadolinia credit to cask transportation of BWR-STEP3 SFAs

    International Nuclear Information System (INIS)

    Kikuchi, Tsukasa; Mitsuhashi, Ishi; Ito, Dai-ichiro; Nakamura, Yu

    2003-01-01

    Instead of the fresh-fuel assumption, the application of gadolinia credit to cask transportation of BWR SFAs is studied. Its efficacy for BWR-STEP2 SFAs had already been estimated. This paper reports on the application of gadolinia credit to cask transportation of BWR-STEP3 SFAs. (author)

  15. Laser cladding technology to small diameter pipes

    International Nuclear Information System (INIS)

    Fujimagari, H.; Hagiwara, M.; Kojima, T.

    2000-01-01

    A laser cladding method which produces a highly corrosion-resistant material coating layers (cladding) on the austenitic stainless steel (type 304 SS) pipe inner surface was developed to prevent SCC (stress corrosion cracking) occurrence. This technology is applicable to a narrow and long distance area from operators, because of the good accessibility of the YAG (yttrium-aluminum-garnet) laser beam that can be transmitted through an optical fiber. In this method a mixed paste metallic powder and heating-resistive organic solvent are firstly placed on the inner surface of a small pipe, and then a YAG laser beam transmitted through an optical fiber irradiates to the pasted area. A mixed paste will be melted and form a cladding layer subsequently. A cladding layer shows as excellent corrosion resistance property. This laser cladding (LC) method had already applied to several domestic nuclear power plants and had obtained a good reputation. This report introduces the outline of laser cladding technology, the developed equipment for practical application in the field, and the circumstance in actual plant application. (orig.)

  16. On-line prediction of BWR transients in support of plant operation and safety analyses

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Lekach, S.V.; Mallen, A.N.

    1983-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology is presented which affords realistic predictions of plant transient and severe off-normal events in LWR power plants through on-line simulations at a speed ten times greater than actual process speeds. Results are shown for a BWR plant simulation. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the recirculation loop and feed-water train. Point kinetics incorporate reactivity feedback for void fraction, for fuel temperature, and for coolant temperature. Control systems and trip logic are simulated for the nuclear steam supply system

  17. Seismic risk assessment of a BWR: status report

    International Nuclear Information System (INIS)

    Chuang, T.Y.; Bernreuter, D.L.; Wells, J.E.; Johnson, J.J.

    1985-02-01

    The seismic risk methodology developed in the US NRC Seismic Safety Margins Research Program (SSMRP) was demonstrated by its application to the Zion nuclear power plant, a pressurized water reactor (PWR). A detailed model of Zion, including systems analysis models (initiating events, event trees, and fault trees), SSI and structure models, and piping models was developed and analyzed. The SSMRP methodology can equally be applied to a boiling water reactor (BWR). To demonstrate its applicability, to identify fundamental differences in seismic risk between a PWR and a BWR, and to provide a basis of comparison of seismic risk between a PWR and a BWR when analyzed with comparable methodology and assumptions, a seismic risk analysis is being performed on the LaSalle County Station nuclear power plant

  18. The BWR owners' group planning guide for life extension

    International Nuclear Information System (INIS)

    Smith, S.K.; Lehnert, D.F.; Locke, R.K.

    1991-01-01

    Extending the operating life of a commercial nuclear power plant has been shown to be economically beneficial to both the utility and the electric customer. As such, many utilities are planning and implementing plant life extension (PLEX) programs. A document has been developed which provides guidance to utilities in formulating a PLEX program plant for one or more boiling water reactor (BWR) plants. The guide has been developed by the BWR Owners' Group Plant Life Extension Committee. The principal bases for this guide were the BWR Pilot and Lead Plant Programs. These programs were used as models to develop the 'base plan' described in this guide. By formulating their program plant utilizing the base plan, utilities will be able to maximize the use of existing evaluations and results. The utility planner will build upon the base plan by adding any tasks or features that are unique to their programs. (author)

  19. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  20. An overview of the BWR ECCS strainer blockage issues

    International Nuclear Information System (INIS)

    Serkiz, A.W.; Marshall, M.L. Jr.; Elliott, R.

    1996-01-01

    This Paper provides a brief overview of actions taken in the mid 1980s to resolve Unresolved Safety Issue (USI) A-43, open-quotes Containment Emergency Sump Performance,close quotes and their relationship to the BWR strainer blockage issue; the importance of insights gained from the Barseback-2 (a Swedish BWR) incident in 1992 and from ECCS strainer testing and inspections at the Perry nuclear power plant in 1992 and 1993; an analysis of an US BWR/4 with a Mark I containment; an international community sharing of knowledge relevant to ECCS strainer blockage, additional experimental programs; and identification of actions needed to resolve the strainer blockage issue and the status of such efforts

  1. FRAPCON analysis of cladding performance during dry storage operations

    Energy Technology Data Exchange (ETDEWEB)

    Richmond, David J.; Geelhood, Kenneth J.

    2018-03-01

    There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

  2. Clad Degradation- Summary and Abstraction for LA

    International Nuclear Information System (INIS)

    D. Stahl

    2004-01-01

    The purpose of this model report is to develop the summary cladding degradation abstraction that will be used in the Total System Performance Assessment for the License Application (TSPA-LA). Most civilian commercial nuclear fuel is encased in Zircaloy cladding. The model addressed in this report is intended to describe the postulated condition of commercial Zircaloy-clad fuel as a function of postclosure time after it is placed in the repository. Earlier total system performance assessments analyzed the waste form as exposed UO 2 , which was available for degradation at the intrinsic dissolution rate. Water in the waste package quickly became saturated with many of the radionuclides, limiting their release rate. In the total system performance assessments for the Viability Assessment and the Site Recommendation, cladding was analyzed as part of the waste form, limiting the amount of fuel available at any time for degradation. The current model is divided into two stages. The first considers predisposal rod failures (most of which occur during reactor operation and associated activities) and postdisposal mechanical failure (from static loading of rocks) as mechanisms for perforating the cladding. Other fuel failure mechanisms including those caused by handling or transportation have been screened out (excluded) or are treated elsewhere. All stainless-steel-clad fuel, which makes up a small percentage of the overall amount of fuel to be stored, is modeled as failed upon placement in the waste packages. The second stage of the degradation model is the splitting of the cladding from the reaction of water or moist air and UO 2 . The splitting has been observed to be rapid in comparison to the total system performance assessment time steps and is modeled to be instantaneous. After the cladding splits, the rind buildup inside the cladding widens the split, increasing the diffusion area from the fuel rind to the waste package interior. This model report summarizes the

  3. Modernization of the oldest Swedish NPP

    International Nuclear Information System (INIS)

    Hagberth, Ronald

    1998-01-01

    OKG operates three BWR units of ABB design: Oskarshamn 1 with a net capacity of 440' MW, Oskarshamn 2 of 600 MW and Oskarshamn 3 of 1160 MW. Oskarshamn 1 NPP was commissioned in 1972 as the first commercial nuclear unit in Sweden. After more than twenty years of successful operation, the unit is now also the first reactor in Sweden to undergo a large safety modernization program. In the year 2000 the Oskarshamn 1 NPP will be modernized to a high level of safety standard and ready for operation for another period of at least 20 years. Experience gained can be used when modernizing other NPPs. The investment program for life extension is reasonable and shows that NPPs can be operated with an expected life span of more than 40 years at an ever-increasing safety level and still be very competitive in a deregulated market. (author)

  4. The JAERI code system for evaluation of BWR ECCS performance

    International Nuclear Information System (INIS)

    Kohsaka, Atsuo; Akimoto, Masayuki; Asahi, Yoshiro; Abe, Kiyoharu; Muramatsu, Ken; Araya, Fumimasa; Sato, Kazuo

    1982-12-01

    Development of respective computer code system of BWR and PWR for evaluation of ECCS has been conducted since 1973 considering the differences of the reactor cooling system, core structure and ECCS. The first version of the BWR code system, of which developmental work started earlier than that of the PWR, has been completed. The BWR code system is designed to provide computational tools to analyze all phases of LOCAs and to evaluate the performance of the ECCS including an ''Evaluation Model (EM)'' feature in compliance with the requirements of the current Japanese Evaluation Guideline of ECCS. The BWR code system could be used for licensing purpose, i.e. for ECCS performance evaluation or audit calculations to cross-examine the methods and results of applicants or vendors. The BWR code system presented in this report comprises several computer codes, each of which analyzes a particular phase of a LOCA or a system blowdown depending on a range of LOCAs, i.e. large and small breaks in a variety of locations in the reactor system. The system includes ALARM-B1, HYDY-B1 and THYDE-B1 for analysis of the system blowdown for various break sizes, THYDE-B-REFLOOD for analysis of the reflood phase and SCORCH-B2 for the calculation of the fuel assembl hot plane temperature. When the multiple codes are used to analyze a broad range of LOCA as stated above, it is very important to evaluate the adequacy and consistency between the codes used to cover an entire break spectrum. The system consistency together with the system performance are discussed for a large commercial BWR. (author)

  5. New method to calculate the mechanical properties of unirradiated fuel cladding from ring tensile tests

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Rengel, M.A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain); Consejo de Seguridad Nuclear (CSN), Justo Dorado 11, E-28040 Madrid (Spain); Gomez, F.J.; Ruiz-Hervias, J.; Caballero, L.; Valiente, A. [Departamento de Ciencia de Materiales, UPM, E.T.S.I. Caminos, Canales y Puertos, Profesor Aranguren s/n, E-28040 Madrid (Spain)

    2009-06-15

    displacement of the loading piece and another one between the equivalent stress in the same point and the nominal applied stress. In the first iteration a calculation is performed with an approximate plastic stress-strain law, and the two above-mentioned relationships are used to determine a new law from the experimental results. In the second iteration the calculation takes into account the new plastic stress-strain law and determines two new relationships. After a few iterations an excellent fit is obtained. This method is an improvement of the original method by Arsene and Bai [3] and allows obtaining the plastic stress-strain curve in the hoop direction in a consistent way. The experimental data used in this work to check the validity of the procedure have been obtained on unirradiated Zirlo cladding, with the standard alloy composition and geometry (outer diameter of the cladding 9.5 mm and a wall thickness of 0.56 mm). References: [1]. Arsene, S.; Bai, J.B. A new approach to measuring transverse properties of structural tubing by a ring test, Journal of Testing and Evaluation, 24: 386-391 (1996) [2]. Arsene, S.; Bai, J. 'A new approach to measuring transverse properties of structural tubing by a ring test-experimental investigation', Journal of Testing and Evaluation, 26: 26-30 (1998) [3]. Arsene, S.; Bai, J.B.; 'Hydride embrittlement and irradiation effects on the hoop mechanical properties of pressurized water reactor (PWR) and boiling-water reactor (BWR) zircaloy cladding tubes: Part I. Hydride embrittlement in stress-relieved, annealed, and recrystallized zircaloys at 20 deg. C and 300 deg. C', Metallurgical and materials and transactions A, 34A: 553-566 (2003) [4]. Chang-Sun Seok, Bong-Kook, K.Linga, 'The properties of the ring and burst creep of zirlo claddings', Engineering Failure Analysis, 13: 389-397 (2006). (authors)

  6. The BWR VIP role in license renewal

    International Nuclear Information System (INIS)

    Dyle, R.

    2001-01-01

    The full text follows. The Boiling Water Reactor Vessel and Internals Project (BWRVIP) was started in response to an increase in the occurrence of intergranular stress corrosion cracking (IGSCC) in BWR reactor pressure vessel (RPV) internal components. The BWRVIP first evaluated the internals to determine which components were necessary to assure safe operation. An assessment of the relative significance of the internals was then performed to establish the priority in which components would be evaluated. Once this was determined, each safety-related component was evaluated to determine what, if any, inspections or tests were necessary to assure component integrity. Although IGSCC was the initial degradation mechanism of concern for RPV internals, the individual component evaluations considered all known modes of failure (fatigue, stress corrosion cracking, neutron embrittlement, etc). The component evaluations also considered all potential failure locations and the susceptibility to degradation. Once the potential failure locations and mechanisms were identified, the BWRVIP developed inspection criteria to assess component condition. The BWRVIP also developed flaw evaluation methodologies that could be used to determine the integrity and remaining life of each component. All of this information was consolidated into a document called an Inspection and Flaw Evaluation (IE) Guideline for each component. At the same time the BWRVIP was developing its program to assure internals integrity, utilities began to seriously consider measures necessary to extend the life of the plants. In the United States, the USNRC promulgated rules to allow the renewal of a license to allow plant operation for an additional 20 years. One aspect of the rule was that management of age-related degradation in the renewal period must be performed. The timing of this ''license renewal rule'' was advantageous in that it allowed the BWRVIP to address the requirements of the rule in the development

  7. Peach Bottom transient analysis with BWR TRACB02

    International Nuclear Information System (INIS)

    Alamgir, M.; Sutherland, W.A.

    1984-01-01

    TRAC calculations have been performed for a Turbine Trip transient (TT1) in the Peach Bottom BWR power plant. This study is a part of the qualification of the BWR-TRAC code. The simulation is aimed at reproducing the observed thermal hydraulic behavior in a pressurization transient. Measured core power is an input to the calculation. Comparison with data show the code reasonably well predicts the generation and propagation of the pressure waves in the main steam line and associated pressurization of the reactor vessel following the closure of the turbine stop valve

  8. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  9. Limerick BWR turbine control and protection system upgrade success

    International Nuclear Information System (INIS)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A.; Williams, J.C.

    2015-01-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  10. Limerick BWR turbine control and protection system upgrade success

    Energy Technology Data Exchange (ETDEWEB)

    Tang, C.K.; Pietryka, T.S.; Federico, P.A., E-mail: tangck@westinghouse.com, E-mail: pietryt@westinghouse, E-mail: federipa@westinghouse.com [Westinghouse Electric Company, LLC, Cranberry Township, PA (United States); Williams, J.C., E-mail: Jonathan.Williams@exeloncorp.com [Exelon Nuclear, Warrenville, IL (United States)

    2015-07-01

    Westinghouse and Exelon have successfully implemented a digital electro-hydraulic control (DEHC) at Limerick BWR Unit 1 Station to perform the turbine control, protection and reactor pressure functions. The DEHC replaces analog controls and addressed system performance, obsolescence and reliability. This was a first-of-a-kind application for control and protection of the main turbine and BWR pressure control for the distributed control system utilized. The demolition of analog equipment, main control room and front standard modifications, and acceptance testing were completed on schedule during the normal 2014 outage. Key aspects of the project that facilitated this success will be discussed and presented. (author)

  11. BWR shutdown analyzer using artificial intelligence (AI) techniques

    International Nuclear Information System (INIS)

    Cain, D.G.

    1986-01-01

    A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

  12. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

  13. BWR stability using a reducing dynamical model; Estabilidad de un BWR con un modelo dinamico reducido

    Energy Technology Data Exchange (ETDEWEB)

    Ballestrin Bolea, J M; Blazquez Martinez, J B

    1990-07-01

    BWR stability can be treated with reduced order dynamical models. When the parameters of the model came from dynamical models. When the parameters of the model came from experimental data, the predictions are accurate. In this work an alternative derivation for the void fraction equation is made, but remarking the physical structure of the parameters. As the poles of power/reactivity transfer function are related with the parameters, the measurement of the poles by other techniques such as noise analysis will lead to the parameters, but the system of equations is non-linear. Simple parametric calculation of decay ratio are performed, showing why BWRs become unstable when they are operated at low flow and high power. (Author)

  14. Fuel rod response to BWR power oscillations during anticipated transient without scram

    International Nuclear Information System (INIS)

    Cunningham, M.; Scott, H.

    1998-01-01

    The US NRC is examining fuel behaviour during a postulated BWR anticipated transient without scram (ATWS) with power oscillations to determine if current regulatory criteria are adequate. Currently, the 280 cal/g limit for RIAs is used to show that coolable geometry is maintained and pressure pulses are avoided during ATWSs. Two specific questions have now been raised about the continued use of the 280 cal/g value. First, this value was derived from energy deposition values whereas the regulatory requirements are written in terms of fuel enthalpy. The second is that fuel rod rupture with fuel dispersal has been observed in RIA tests with high bum-up fuel rods having energy deposition values well below the current limit. However, the BWR ATWS power oscillation transient is slower than a RIA power pulse, thus reducing the likelihood of failure. Therefore questions about the adequacy of the 280 cal/g limit do not necessarily imply unacceptable fuel damage occurring during such power oscillations and there is no immediate safety concern. The reported analysis, using the FRAPTRAN transient fuel rod analysis code, was thus undertaken to determine if further investigation might be appropriate and with the intention of starting some discussions about the issue. There was a comment that a limit of 100 cal/g fuel enthalpy had been mentioned following the scoping calculations but that perhaps enthalpy was not the main concern in an ATWS. It was also observed that cladding stresses are lower than in all RIA. The question was what really is the main concern. It was replied that the main concern was a question of maintaining a coolable geometry i.e. not loosing fuel particles out of the rod. And it was agreed that enthalpy may not be the important issue, rather that it previously had been used as the parameter and so had been considered. Confirmation of this presently being an evaluation and not a regulatory concern was sought and provided, it being pointed out that the NRC

  15. A pellet-clad interaction failure criterion

    International Nuclear Information System (INIS)

    Howl, D.A.; Coucill, D.N.; Marechal, A.J.C.

    1983-01-01

    A Pellet-Clad Interaction (PCI) failure criterion, enabling the number of fuel rod failures in a reactor core to be determined for a variety of normal and fault conditions, is required for safety analysis. The criterion currently being used for the safety analysis of the Pressurized Water Reactor planned for Sizewell in the UK is defined and justified in this paper. The criterion is based upon a threshold clad stress which diminishes with increasing fast neutron dose. This concept is consistent with the mechanism of clad failure being stress corrosion cracking (SCC); providing excess corrodant is always present, the dominant parameter determining the propagation of SCC defects is stress. In applying the criterion, the SLEUTH-SEER 77 fuel performance computer code is used to calculate the peak clad stress, allowing for concentrations due to pellet hourglassing and the effect of radial cracks in the fuel. The method has been validated by analysis of PCI failures in various in-reactor experiments, particularly in the well-characterised power ramp tests in the Steam Generating Heavy Water Reactor (SGHWR) at Winfrith. It is also in accord with out-of-reactor tests with iodine and irradiated Zircaloy clad, such as those carried out at Kjeller in Norway. (author)

  16. Fuel assembly and fuel cladding tube

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

    1996-01-01

    A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

  17. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Jen, C.-K.; Ihara, I.

    1999-01-01

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  18. Inservice inspection of Halden BWR pressure vessel

    International Nuclear Information System (INIS)

    Foerli, O.; Hernes, T.

    1978-01-01

    A description is given of how the recertification inspection of the 20 years old Halden Reactor pressure vessel was carried out in accordance with the latest ASME-CODES, despite the fact that inspection accessibility was poor. As no volumetric inspection had been carried out since the preservice radiography in 1957, the ultrasonic inspection included the high flux region of all welds. In total 70% of longitudinal welds and 20% of bottom circumferential welds were inspected as well as the bottom nozzle connection. The vessel was not designed with provisions for inservice inspection, the welds are unaccessible from the outside and removal of the lid is virtually impossible. The ultrasonic probes could only be loaded through 77 mm diameter holes in the top lid and remotely positioned inside the vessel. The inspection was performed using 450C and 60OC 1 MHz angle probes and 2.25 MHz normal probes in immersion technique. In a zone around the welds, small regions with lack of bonding between the stainless steel cladding and the boiler steel were revealed. One root defect known and accepted from the preservice radiographs was examined. The defect was found to be 6x30mm as a maximum and well within acceptable limits according to the fracture mechanics analysis method recommended in ASME X1. The inspection required a period of three weeks' work in the reactor hall. (UK)

  19. Evaluation of PWR and BWR pin cell benchmark results

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J.; Hoogenboom, J.E.; Leege, P.F.A. de; Voet, J. van der; Verhagen, F.C.M.

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs

  20. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pijlgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs., 9 figs., 30 tabs.

  1. Requests on domestic nuclear data library from BWR design

    International Nuclear Information System (INIS)

    Maruyama, Hiromi

    2003-01-01

    Requests on the domestic nuclear data library JENDL and activities of the Nuclear Data Center have been presented from the perspective of BWR design and design code development. The requests include a standard multi-group cross section library, technical supports, and clarification of advantage of JENDL as well as requests from physical aspects. (author)

  2. BWR ATWS mitigation by Fine Motion Control Rod

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.; Mallen, A.; Diamond, D.

    1994-01-01

    Two main methods of ATWS mitigation in a SBWR are: fine Motion control Rods (FMCRD) and Boron injection via the Standby Liquid control System (SLCS). This study has demonstrated that the use of FMCRD along with feedwater runback mitigated the conditions due to reactivity insertion and possible ATWS in a BWR which is similar to SBWR

  3. Detection of failed fuel rods in shrouded BWR fuel assemblies

    International Nuclear Information System (INIS)

    Baero, G.; Boehm, W.; Goor, B.; Donnelly, T.

    1988-01-01

    A manipulator and an ultrasonic testing (UT) technique were developed to identify defective fuel rods in shrouded BWR fuel assemblies. The manipulator drives a UT probe axially through the bottom tie plate into the water channels between the fuel rods. The rotating UT probe locates defective fuel rods by ingressed water which attenuates the UT-signal. (author)

  4. Transmutation of minor actinide using thorium fueled BWR core

    International Nuclear Information System (INIS)

    Susilo, Jati

    2002-01-01

    One of the methods to conduct transmutation of minor actinide is the use of BWR with thorium fuel. Thorium fuel has a specific behaviour of producing a little secondary minor actinides. Transmutation of minor actinide is done by loading it in the BWR with thorium fuel through two methods, namely close recycle and accumulation recycle. The calculation of minor actinide composition produced, weigh of minor actinide transmuted, and percentage of reminder transmutation was carried SRAC. The calculations were done to equivalent cell modeling from one fuel rod of BWR. The results show that minor actinide transmutation is more effective using thorium fuel than uranium fuel, through both close recycle and accumulation recycle. Minor actinide transmutation weight show that the same value for those recycle for 5th recycle. And most of all minor actinide produced from 5 unit BWR uranium fuel can transmuted in the 6 t h of close recycle. And, the minimal value of excess reactivity of the core is 12,15 % Δk/k, that is possible value for core operation

  5. Physical model of nonlinear noise with application to BWR stability

    International Nuclear Information System (INIS)

    March-Leuba, J.; Perez, R.B.

    1983-01-01

    Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD

  6. Power plant design: ESBWR - the latest passive BWR

    International Nuclear Information System (INIS)

    Arnold, H.; Yadigaroglu, G.; Stoop, P.C.

    1997-01-01

    When General Electric said it would end development of its 670 MWe SBWR (Simplified Boiling Water Reactor), it was not quite the end of the story. Also on the drawing board at the time was the larger ESBWR (standing for either European or Economic Simplified BWR) whose goal was to provide the improved economic performance that the SBWR could not. (UK)

  7. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

    2008-07-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  8. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

    2008-01-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  9. Decay ratio studies in BWR and PWR using wavelet

    International Nuclear Information System (INIS)

    Ciftcioglu, Oe.

    1996-10-01

    The on-line stability of BWR and PWR is studied using the neutron noise signals as the fluctuations reflect the dynamic characteristics of the reactor. Using appropriate signal modeling for time domain analysis of noise signals, the stability parameters can be directly obtained from the system impulse response. Here in particular for BWR, an important stability parameter is the decay ratio (DR) of the impulse response. The time series analysis involves the autoregressive modeling of the neutron detector signal. The DR determination is strongly effected by the low frequency behaviour since the transfer function characteristic tends to be a third order system rather than a second order system for a BWR. In a PWR low frequency behaviour is modified by the Boron concentration. As a result of these phenomena there are difficulties in the consistent determination of the DR oscillations. The enhancement of the consistency of this DR estimation is obtained by wavelet transform using actual power plant data from BWR and PWR. A comparative study of the Restimation with and without wavelets are presented. (orig.)

  10. Operator training simulator for BWR nuclear power plant

    International Nuclear Information System (INIS)

    Watanabe, Tadasu

    1988-01-01

    For the operation management of nuclear power stations with high reliability and safety, the role played by operators is very important. The effort of improving the man-machine interface in the central control rooms of nuclear power stations is energetically advanced, but the importance of the role of operators does not change. For the training of the operators of nuclear power stations, simulators have been used from the early stage. As the simulator facilities for operator training, there are the full scope simulator simulating faithfully the central control room of an actual plant and the small simulator mainly aiming at learning the plant functions. For BWR nuclear power stations, two full scope simulators are installed in the BWR Operator Training Center, and the training has been carried out since 1974. The plant function learning simulators have been installed in respective electric power companies as the education and training facilities in the companies. The role of simulators in operator training, the BTC No.1 simulator of a BWR-4 of 780 MWe and the BTC No.2 simulator of a BWR-5 of 1,100 MWe, plant function learning simulators, and the design of the BTC No.2 simulator and plant function learning simulators are reported. (K.I.)

  11. Capturing reflected cladding modes from a fiber Bragg grating with a double-clad fiber coupler.

    Science.gov (United States)

    Baiad, Mohamad Diaa; Gagné, Mathieu; Lemire-Renaud, Simon; De Montigny, Etienne; Madore, Wendy-Julie; Godbout, Nicolas; Boudoux, Caroline; Kashyap, Raman

    2013-03-25

    We present a novel measurement scheme using a double-clad fiber coupler (DCFC) and a fiber Bragg grating (FBG) to resolve cladding modes. Direct measurement of the optical spectra and power in the cladding modes is obtained through the use of a specially designed DCFC spliced to a highly reflective FBG written into slightly etched standard photosensitive single mode fiber to match the inner cladding diameter of the DCFC. The DCFC is made by tapering and fusing two double-clad fibers (DCF) together. The device is capable of capturing backward propagating low and high order cladding modes simply and efficiently. Also, we demonstrate the capability of such a device to measure the surrounding refractive index (SRI) with an extremely high sensitivity of 69.769 ± 0.035 μW/RIU and a resolution of 1.433 × 10(-5) ± 8 × 10(-9) RIU between 1.37 and 1.45 RIU. The device provides a large SRI operating range from 1.30 to 1.45 RIU with sufficient discrimination for all individual captured cladding modes. The proposed scheme can be adapted to many different types of bend, temperature, refractive index and other evanescent wave based sensors.

  12. Inspection system for Zircaloy clad fuel rods

    International Nuclear Information System (INIS)

    Yancey, M.E.; Porter, E.H.; Hansen, H.R.

    1975-10-01

    A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory (ANL). Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers (LVDTs) and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods

  13. Fuel cladding mechanical properties for transient analysis

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

    1976-01-01

    Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

  14. Laser cladding to select new glassy alloys

    International Nuclear Information System (INIS)

    Medrano, L.L.O.; Afonso, C.R.M.; Kiminami, C.S.; Gargarella, P.; Ramasco, B.

    2016-01-01

    A new experimental technique used to analyze the effect of compositional variation and cooling rate in the phase formation in a multicomponent system is the laser cladding. This work have evaluated the use of laser cladding to discover a new bulk metallic glass (BMG) in the Al-Co-Zr system. Coatings with composition variation have made by laser cladding using Al-Co-Zr alloys powders and the samples produced have been characterized by X ray diffraction, microscopy and energy-dispersive X-ray spectroscopy. The results did not show the composition variation as expected, because of incomplete melting during laser process. It was measured a composition variation tendency that allowed the glass forming investigation by the glass formation criterion λ+Δh 1/2 . The results have showed no glass formation in the coating samples, which prove a limited capacity of Zr-Co-Al system to form glass (author)

  15. Potential effects of gallium on cladding materials

    International Nuclear Information System (INIS)

    Wilson, D.F.; Beahm, E.C.; Besmann, T.M.; DeVan, J.H.; DiStefano, J.R.; Gat, U.; Greene, S.R.; Rittenhouse, P.L.; Worley, B.A.

    1997-10-01

    This paper identifies and examines issues concerning the incorporation of gallium in weapons derived plutonium in light water reactor (LWR) MOX fuels. Particular attention is given to the more likely effects of the gallium on the behavior of the cladding material. The chemistry of weapons grade (WG) MOX, including possible consequences of gallium within plutonium agglomerates, was assessed. Based on the calculated oxidation potentials of MOX fuel, the effect that gallium may have on reactions involving fission products and possible impact on cladding performance were postulated. Gallium transport mechanisms are discussed. With an understanding of oxidation potentials and assumptions of mechanisms for gallium transport, possible effects of gallium on corrosion of cladding were evaluated. Potential and unresolved issues and suggested research and development (R and D) required to provide missing information are presented

  16. Stress corrosion testing of irradiated cladding tubes

    International Nuclear Information System (INIS)

    Lunde, L.; Olshausen, K.D.

    1980-01-01

    Samples from two fuel rods with different cladding have been stress corrosion tested by closed-end argon-iodine pressurization at 320 0 C. The fuel rods with stress relieved and recrystallized Zircaloy-2 had received burnups of 10.000 and 20.000 MWd/ton UO 2 , respectively. It was found that the SCC failure stress was unchanged or slightly higher for the irradiated than for the unirradiated control tubes. The tubes failed consistently in the end with the lowest irradiation dose. The diameter increase of the irradiated cladding during the test was 1.1% for the stress-relieved samples and 0.24% for the recrystallized samples. SEM examination revealed no major differences between irradiated and unirradiated cladding. A ''semi-ductile'' fracture zone in recrystallized material is described in some detail. (author)

  17. Assessment of the Prony's method for BWR stability analysis

    International Nuclear Information System (INIS)

    Ortiz-Villafuerte, Javier; Castillo-Duran, Rogelio; Palacios-Hernandez, Javier C.

    2011-01-01

    Highlights: → This paper describes a method to determine the degree of stability of a BWR. → Performance comparison between Prony's and common AR techniques is presented. → Benchmark data and actual BWR transient data are used for comparison. → DR and f results are presented and discussed. → The Prony's method is shown to be a robust technique for BWR stability. - Abstract: It is known that Boiling Water Reactors are susceptible to present power oscillations in regions of high power and low coolant flow, in the power-flow operational map. It is possible to fall in one of such instability regions during reactor startup, since both power and coolant flow are being increased but not proportionally. One other possibility for falling into those areas is the occurrence of a trip of recirculation pumps. Stability monitoring in such cases can be difficult, because the amount or quality of power signal data required for calculation of the stability key parameters may not be enough to provide reliable results in an adequate time range. In this work, the Prony's Method is presented as one complementary alternative to determine the degree of stability of a BWR, through time series data. This analysis method can provide information about decay ratio and oscillation frequency from power signals obtained during transient events. However, so far not many applications in Boiling Water Reactors operation have been reported and supported to establish the scope of using such analysis for actual transient events. This work presents first a comparison of decay ratio and frequency oscillation results obtained by Prony's method and those results obtained by the participants of the Forsmark 1 and 2 Boiling Water Reactor Stability Benchmark using diverse techniques. Then, a comparison of decay ratio and frequency oscillation results is performed for four real BWR transient event data, using Prony's method and two other techniques based on an autoregressive modeling. The four

  18. Comparison of corrosion behavior between fusion cladded and explosive cladded Inconel 625/plain carbon steel bimetal plates

    International Nuclear Information System (INIS)

    Zareie Rajani, H.R.; Akbari Mousavi, S.A.A.; Madani Sani, F.

    2013-01-01

    Highlights: ► Both explosive and fusion cladding aggravate the corrosion resistance of Inconel 625. ► Fusion cladding is more detrimental to nonuniform corrosion resistance. ► Single-layered fusion coat does not show any repassivation ability. ► Adding more layers enhance the corrosion resistance of fusion cladding Inconel 625. ► High impact energy spoils the corrosion resistance of explosive cladding Inconel 625. -- Abstract: One of the main concerns in cladding Inconel 625 superalloy on desired substrates is deterioration of corrosion resistance due to cladding process. The present study aims to compare the effect of fusion cladding and explosive cladding procedures on corrosion behavior of Inconel 625 cladding on plain carbon steel as substrate. Also, an attempt has been made to investigate the role of load ratio and numbers of fusion layers in corrosion behavior of explosive and fusion cladding Inconel 625 respectively. In all cases, the cyclic polarization as an electrochemical method has been applied to assess the corrosion behavior. According to the obtained results, both cladding methods aggravate the corrosion resistance of Inconel 625. However, the fusion cladding process is more detrimental to nonuniform corrosion resistance, where the chemical nonuniformity of fusion cladding superalloy issuing from microsegregation, development of secondary phases and contamination of clad through dilution hinders formation of a stable passive layer. Moreover, it is observed that adding more fusion layers can enhance the nonuniform corrosion resistance of fusion cladding Inconel 625, though this resistance still remains weaker than explosive cladding superalloy. Also, the results indicate that raising the impact energy in explosive cladding procedure drops the corrosion resistance of Inconel 625.

  19. Modelling of pellet-clad interaction during power ramps

    International Nuclear Information System (INIS)

    Zhou, G.; Lindback, J.E.; Schutte, H.C.; Jernkvist, L.O.; Massih, A.R.; Massih, A.R.

    2005-01-01

    A computational method to describe the pellet-clad interaction phenomenon is presented. The method accounts for the mechanical contact between fragmented pellets and the zircaloy clad, as well as for chemical reaction of fission products with zircaloy during power ramps. Possible pellet-clad contact states, soft, hard and friction, are taken into account in the computational algorithm. The clad is treated as an elastic-plastic-viscoplastic material with irradiation hardening. Iodine-induced stress corrosion cracking is described by using a fracture mechanics-based model for crack propagation. This integrated approach is used to evaluate two power ramp experiments made on boiling water reactor fuel rods in test reactors. The influence of the pellet-clad coefficient of friction on clad deformation is evaluated and discussed. Also, clad deformations, pellet-clad gap size and fission product gas release for one of the ramped rods are calculated and compared with measured data. (authors)

  20. Multilayer cladding with hyperbolic dispersion for plasmonic waveguides

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia; Shalaginov, Mikhail Y.; Ishii, Satoshi

    2015-01-01

    We study the properties of plasmonic waveguides with a dielectric core and multilayer metal-dielectric claddings that possess hyperbolic dispersion. The waveguides hyperbolic multilayer claddings show better performance in comparison to conventional plasmonic waveguides. © OSA 2015....

  1. Cladding using a 15 kW CO2 laser

    International Nuclear Information System (INIS)

    Vesely, E.J.; Verma, S.K.

    1989-01-01

    Laser alloying or cladding differs little in principle from the traditional forms of weld overlays, but lasers as a heat source offer some distinct advantages. With the selective heating attainable using high power lasers, good metallurgical bond of the clad layer, minimal dilution and typically, a very fine homogeneous microstructure can be obtained in the clad layer. This is a review of work in laser cladding using the 15 kW CO 2 laser. The authors discuss the ability of the laser clad surface to increase the high temperature oxidation resistance of a low-alloy carbon steel (4140). Examples of clads subjected to high- temperature thermal cycling of nickel-20% aluminum and TaC + 4140 clad low-alloy steel and straight high-temperature oxidation of Stellite 6-304L cladding on a 4140 substrate are given

  2. Evaluation of the thermal-mechanical performance of fuel rods of a BWR during a power ramp using the FUELSIM code

    International Nuclear Information System (INIS)

    Pantoja C, R.

    2010-01-01

    To avoid the risk to environment due to release of radioactive material, because of occurrence of an accident, it is the priority of the design and performance of the diverse systems of safety of a commercial nuclear power plant. The safety of nuclear power plants requires, therefore, monitoring those parameters having some direct or indirect effect on safety. The thermal limits are values set for those parameters considered having most impact on the safe operation of a nuclear power reactor. Some thermal limits monitoring requires the thermal-mechanical analysis of the rods containing the nuclear fuel. The fuel rod thermal-mechanical behavior under irradiation is a complex process in which there exists a great deal of interrelated physical and chemical phenomena, so that the fuel rod performance analysis in the core of a nuclear power reactor is generally accomplished by using computer codes, which integrate several of the phenomena that are expected to occur during the lifetime of the fuel rod in the core. The main application of the thermal-mechanical analysis codes is the prediction of occurrence of conditions and/or phenomena that could lead to the deterioration or even mechanical failure of the fuel rod cladding, as, for example, the pellet-cladding interaction. In the operation of a nuclear power reactor, fuel preconditioning operations refer to the operational procedures employed to reduce the fuel rod failure probability due to fuel-cladding interaction, specially during reactor startup. Preconditioning simulations are therefore necessary to determine in advance limit values for the power that can be generated in a fuel rod, and thus avoiding any rod damage. In this work, a first analysis of the thermal-mechanical performance of typical fuel rods used in nuclear reactors of the type BWR 5/6, as those two nuclear reactors in Laguna Verde, Veracruz, is performed. This study includes two types of fuel rods: one from a fuel assembly design with an array 8 x 8

  3. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  4. Metallurgical factors that contribute to cracking in BWR piping

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    During the fall of 1974 and early winter of 1975, cracks have been discovered in the 4 in. bypass lines of several Boiling Water Reactors (BWR's) in the United States. Further, similar cracks were discovered at two BWR's in Japan during the same period. More recently, cracks have been discovered in the core spray piping and in a furnace-sensitized ''safe end'' and adjacent ''dutchman'' at the Dresden Nuclear Power Station, Unit No. 2. Although inspections at all other U.S. BWR's have not disclosed further instances of cracking in core spray piping, leaking cracks have been found in the core spray piping of two BWR's overseas. Metallurgical examinations of these cracks are not yet complete. The following observations have been made to date. All cracks (except those in the furnace-sensitized safe end and dutchman) occurred in seamless type 304 stainless steel piping or in elbows fabricated from such piping, in the outer heat affected zone of either field or shop welds, in lines isolated from the main primary coolant flow during full power operation, except for the not yet examined cracks in the Monticello bypass lines. The cracks are exclusively intergranular, and occur in metal that has been lightly sensitized by the welding process, with only intermittent grain boundary carbides. They developed in the areas of peak axial residual stresses from welding rather than in the most heavily sensitized areas. No fatigue striations have been found on the fracture surfaces. The evidence received to date strongly indicates that these cracks were caused by intergranular stress corrosion of weld-sensitized stainless steel by BWR water containing greater than 0.2 ppM oxygen. The possible role of fatigue or alternating stresses in this corrosion is not clear. Further, not all the cracks detected to date necessarily have occurred by the same mechanism

  5. Evaluation of fast experimental reactor claddings, (2)

    International Nuclear Information System (INIS)

    Miura, Makoto; Nagaki, Hiroshi; Koyama, Masahiro; Tanaka, Yasumasa

    1974-01-01

    Thin-walled fine tubes of Type 316 austenitic stainless steel are used for fuel cladding in Joyo (experimental FBR). The material exhibits the change of the mechanical properties in long-time annealing at high temperature, resulting from the precipitation of carbide in structure. In this connection, the experiment and the results on the changes of the microstructure and mechanical properties (proof stress and hardness) are described. The test specimens are the fuel cladding tubes produced for trial for Joyo core and those for FFTF core made in the U.S.A. They were heated between 400 0 and 850 0 C for 1000 hr in vacuum. (Mori, K.)

  6. Management of cladding hulls and fuel hardware

    International Nuclear Information System (INIS)

    1985-01-01

    The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present

  7. Inpile (in PWR) testing of cladding materials

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    As an introduction, the reasons to perform inpile tests are depicted. An overview over general inpile test procedure is given, and test details which are necessary for the development of new alloys for high burnup claddings, like sample geometries and measuring techniques for inpile corrosion testing, are described in detail. Tests for the creep and length change behavior of cladding tubes are described briefly. Finally, conclusions are drawn and literature citations for further test details are given. (author). 9 refs., 2 tabs., 17 figs

  8. Microstructure of laser cladded martensitic stainless steel

    CSIR Research Space (South Africa)

    Van Rooyen, C

    2006-08-01

    Full Text Available and martensite with 10% ferrite for Material B. Table 7 - Proposed martensitic stainless steel alloys for laser cladding Material C* Cr Ni Mn Si Mo Co Ms (ºC)* Cr eq Ni eq Material A 0.4 13 - 1 0.5 2.5 5.5 120 16.5 12.5 Material B 0.2 15 2 1 0.7 2.5 5.5 117... dilution, low heat input, less distortion, increased mechanical and corrosion properties excellent repeatability and control of process parameters. Solidification of laser cladded martensitic stainless steel is primarily austenitic. Microstructures...

  9. Phosphate-core silica-clad Er/Yb-doped optical fiber and cladding pumped laser.

    Science.gov (United States)

    Egorova, O N; Semjonov, S L; Velmiskin, V V; Yatsenko, Yu P; Sverchkov, S E; Galagan, B I; Denker, B I; Dianov, E M

    2014-04-07

    We present a composite optical fiber with a Er/Yb co-doped phosphate-glass core in a silica glass cladding as well as cladding pumped laser. The fabrication process, optical properties, and lasing parameters are described. The slope efficiency under 980 nm cladding pumping reached 39% with respect to the absorbed pump power and 28% with respect to the coupled pump power. Due to high doping level of the phosphate core optimal length was several times shorter than that of silica core fibers.

  10. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  11. Electron beam cladding of titanium on stainless steel plate

    International Nuclear Information System (INIS)

    Tomie, Michio; Abe, Nobuyuki; Yamada, Masanori; Noguchi, Shuichi.

    1990-01-01

    Fundamental characteristics of electron beam cladding was investigated. Titanium foil of 0.2mm thickness was cladded on stainless steel plate of 3mm thickness by scanning electron beam. Surface roughness and cladded layer were analyzed by surface roughness tester, microscope, scanning electron microscope and electron probe micro analyzer. Electron beam conditions were discussed for these fundamental characteristics. It is found that the energy density of the electron beam is one of the most important factor for cladding. (author)

  12. Modernity: Are Modern Times Different?

    Directory of Open Access Journals (Sweden)

    Lynn Hunt

    2014-12-01

    Full Text Available “Modernity” has recently been the subject of considerable discussion among historians. This article reviews some of the debates and argues that modernity is a problematic concept because it implies a complete rupture with “traditional” ways of life. Studies of key terms are undertaken with the aid of Google Ngrams. These show that “modernity,” “modern times,” and “traditional” —in English and other languages— have a history of their own. A brief analysis of the shift from a self oriented toward equilibrium to a self oriented toward stimulation demonstrates that modernity is not necessary to historical analysis.

  13. Polarization effects in silicon-clad optical waveguides

    Science.gov (United States)

    Carson, R. F.; Batchman, T. E.

    1984-01-01

    By changing the thickness of a semiconductor cladding layer deposited on a planar dielectric waveguide, the TE or TM propagating modes may be selectively attenuated. This polarization effect is due to the periodic coupling between the lossless propagating modes of the dielectric slab waveguide and the lossy modes of the cladding layer. Experimental tests involving silicon claddings show high selectivity for either polarization.

  14. Experimental assessment of fuel-cladding interactions

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Elizabeth Sooby [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-29

    A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.

  15. The measurement of residual stresses in claddings

    International Nuclear Information System (INIS)

    Hofer, G.; Bender, N.

    1978-01-01

    The ring core method, a variation of the hole drilling method for the measurement of biaxial residual stresses, has been extended to measure stresses from depths of about 5 to 25mm. It is now possible to measure the stress profiles of clad material. Examples of measured stress profiles are shown and compared with those obtained with a sectioning technique. (author)

  16. Cladding For Transversely-Pumped Laser Rod

    Science.gov (United States)

    Byer, Robert L.; Fan, Tso Yee

    1989-01-01

    Combination of suitable dimensioning and cladding of neodymium:yttrium aluminum garnet of similar solid-state laser provides for more efficient utilization of transversely-incident pump light from diode lasers. New design overcomes some of limitations of longitudinal- and older transverse-pumping concepts and promotes operation at higher output powers in TEM00 mode.

  17. Prevention of nuclear fuel cladding materials corrosion

    International Nuclear Information System (INIS)

    Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K.

    1983-01-01

    The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. (Author)

  18. Mechanism for iodine cracking of zirconium claddings

    International Nuclear Information System (INIS)

    Novikov, V.V.

    1991-01-01

    The mechanism of iodine cracking of zirconium cladding is analyzed taking into account the effect of stresses on diffusion. A decisive effect of the stress gradiemt on crack propagation in an agressive medium is shown. The experimental data are compared with the proposed model

  19. Thermodynamics of pellet-cladding interaction

    International Nuclear Information System (INIS)

    Kyoh, Bunkei; Fuji, Kensho

    1987-01-01

    Equilibrium thermodynamic calculations are performed on the U-Zr-Cs-I-O system that is assumed to exist in the fuel-cladding gap of light water reactor (LWR) fuel under pellet-cladding interaction (PCI) failure condition. For this purpose a computer program called SOLGASMIX-PV for the calculation of complex multi-component equilibria is used, and the results of postirradiation examination are interpreted. The analysis of the thermodynamics of the system U-Zr-Cs-I-O indicates that cesium and iodine are assumed to be released from fuel pellet into the fuel-cladding gap as CsI, therefore, the Cs/I ratio in fuel-cladding bonding zone is one. The important condensed phases in this region are UO 2 , U 3 O 8 , Cs 2 U 2 O 7 , Cs 2 U 15 O 46 , ZrO 2 and CsI, and the major gaseous species are CsI, I 2 and I. Under this situation where Cs/I ratio is one, cesium-zirconate is not present. If, however, cesium rich phase is partially present then cesium will be associated with zirconium, possibly as Cs 2 ZrO 3 . (author)

  20. Study and Behaviour of Prefabricated Composite Cladding

    Science.gov (United States)

    Sai Avinash, P.; Thiagarajan, N.; Santhi, A. S.

    2017-07-01

    The incessant population rise entailed for an expeditious construction at competitive prices that steered the customary path to the light weight structural components. This lead to construction of structural components using ferrocement. The load bearing structural cladding, sizing 3200x900x100 mm, is chosen for the study, which, is analyzed using the software ABAQUS 6.14 in accordance with the IS:875-87 Part1, IS:875-87 Part2, ACI 549R-97, ACI 318R-08 and NZS:3101-06 Part1 standards. The Ferrocement claddings (FCs) are fabricated to a scaled dimension of 400x115x38 mm. The light weight-high strength phenomena are corroborated by incorporating Glass Fibre Reinforced Polymer Laminates (GFRPL) of thickness 6mm, engineered with the aid of hand layup (wet layup) technique wielding epoxy resin, followed by curing under room temperature. The epoxy resin is employed for fastening ferrocement cladding with the Glass fiber reinforced polymer laminate, with the contemporary methodology. The compressive load carrying capacity of the amalgamated assembly, both in presence and absence of Glass Fibre Reinforced polymer laminates (GFRPL) on either side of Ferrocement cladding, has been experimented.

  1. Method for decontaminating stainless cladding tubes

    International Nuclear Information System (INIS)

    Komatsu, Fumiaki.

    1986-01-01

    Purpose: To form an oxide film over the surface of stainless cladding tubes and to efficiently remove radioactive materials from the steel surface together with the oxide layer by the use of an acid water solution. Method: After the removal of water from cladding tubes that have passed through the re-processing process, an oxide film is formed on the surface of the cladding tubes by heating over 400 deg C in an oxidizing atmosphere and thereafter washed again in an acid water solution. When the cladding tubes are thus oxidized once, the stainless base metal itself is oxidized, an oxide layer of several 10 μm or more being formed thereon. In consequence, since the oxide layer is far inferior in corrosion resistance to stainless metals, a pickling liquid easily penetrates into the stainless metal through the oxide layer, thereby remarkably promoting the peeling of the layer from the base metal surface and also improving the residual radioactive material removing efficiency together. (Takahashi, M.)

  2. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    2000-01-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  3. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Sattari-Far, Iradi; Andersson, Magnus

    2006-06-01

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the measurement of

  4. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the

  5. 3D pin-by-pin power density profiles with high spatial resolution in the vicinity of a BWR control blade tip simulated with coupled neutronics/burn-up calculations

    International Nuclear Information System (INIS)

    Li, J.; Nünighoff, K.; Allelein, H.-J.

    2011-01-01

    Highlights: ► High spatial resolution neutronic and burn-up calculations of quarter BWR fuel element section. ► Coupled MCNP(X)–ORIGEN2.2 simulation using VESTA. ► Control blade history effect was taken into account. ► Determining local power excursion after instantaneous control rod movement. ► Correlation between control blade geometry and occurrence of local power excursions. - Abstract: Pellet cladding interaction (PCI) as well as pellet cladding mechanical interaction (PCMI) are well-known fuel failures in light water reactors, especially in boiling water reactors (BWR). Whereas the thermo-mechanical processes of PCI effects have been intensively investigated in the last decades, only rare information is available on the role of neutron physics. However, each power transient is primary due to neutron physics effects and thus knowledge of the neutron physical background is mandatory to better understand the occurrence of PCI effects in BWRs. This paper will focus on a study of local power excursions in a typical BWR fuel assembly during control rod movements. Burn-up and energy deposition were simulated with high spatial granularity, especially in the vicinity of the control blade tip. It could be shown, that the design of the control blade plays a dominant role for the occurrence of local power peaks while instantaneously moving down the control rod. The main result is, that the largest power peak occurs at the interface between steel handle and absorber rods. A full width half maximum (FWHM) of ±2.5 cm was observed. This means, the local power excursion due to neutron physics phenomena involve approximately five pellets. With the VESTA code coupled MCNP(X)/ORIGEN2.2 calculations were performed with more than 3400 burn-up zones in order to take history effects into account.

  6. LAPUR5 BWR stability analysis in Kuosheng nuclear power plant

    International Nuclear Information System (INIS)

    Kunlung Wu; Chunkuan Shih; Wang, J.R.; Kao, L.S.

    2005-01-01

    Full text of publication follows: Unstable oscillation of a nuclear power reactor core is one of the main reasons that causes minor core damage. Stability analysis needs to be performed to predict the potential problem as early as possible and to prevent core instability events from happening. Nuclear Regulatory Commission (NRC) requests all BWR licensees to examine each core reload and to impose operating limitations, as appropriate, to ensure compliance with GDC 10 and 12. GDC 10 requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits will not be exceeded during any condition of normal operation, including anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. Therefore, the core instability is directly related to the fuel design limits. The core and channel DR (decay ratio) calculation are commonly performed to determine system's stability when new fuel designs are introduced in the core. In order to establish the independent analysis technology for BWR licensees and verifications, the Institute of Nuclear Energy Research (INER) has obtained agreement from NRC and implemented the 'Methodology and Procedure for Calculation of Core and Channel Decay Ratios with LAPUR', which was developed by the IBERINCO in 2001. LAPUR5 uses a multi-nodal description of the neutron dynamics, together with a distributed parameter model of the core thermal hydrodynamics to produce a space-dependent representation of the dynamics of a BWR in the frequency domain for small perturbations around a steady state condition. From the output of LAPUR5, the following results are obtained: global core decay ratio, out-of phase core decay ratio, and channel decay ratio. They are key parameters in the determination of BWR core stability

  7. Evolution of transmission spectra of double cladding fiber during etching

    Science.gov (United States)

    Ivanov, Oleg V.; Tian, Fei; Du, Henry

    2017-11-01

    We investigate the evolution of optical transmission through a double cladding fiber-optic structure during etching. The structure is formed by a section of SM630 fiber with inner depressed cladding between standard SMF-28 fibers. Its transmission spectrum exhibits two resonance dips at wavelengths where two cladding modes have almost equal propagation constants. We measure transmission spectra with decreasing thickness of the cladding and show that the resonance dips shift to shorter wavelengths, while new dips of lower order modes appear from long wavelength side. We calculate propagation constants of cladding modes and resonance wavelengths, which we compare with the experiment.

  8. Method for automatic filling of nuclear fuel rod cladding tubes

    International Nuclear Information System (INIS)

    Bezold, H.

    1979-01-01

    Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. (DG) [de

  9. Impact of advanced BWR core physics method on BWR core monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Moon, H; Wells, A [Siemens Power Corporation, Richland (United States)

    2000-07-01

    Siemens Power Corporation recently initiated development of POWERPLEX{sup TM}-III for delivery to the Grand Gulf Nuclear Power Station. The main change introduced in POWERPLEX{sup TM}-III as compared to its predecessor POWERPLEX{sup TM}-II is the incorporation of the advances BWR core simulator MICROBURN-B2. A number of issues were identified and evaluated relating to the implementation of MICROBURN-B2 and its impact on core monitoring. MICROBURN-B2 demands about three to five times more memory and two to three times more computing time than its predecessor MICROBURN-B in POWERPLEX {sup TM}-II. POWERPLEX{sup TM}-III will improve thermal margin prediction accuracy and provide more accurate plant operating conditions to operators than POWERPLEX{sup TM}-II due to its improved accuracy in predicted TIP values and critical k-effective. The most significant advantage of POWERPLEX{sup TM}-III is its capability to monitor a relaxed rod sequence exchange operation. (authors)

  10. Modern cosmology

    International Nuclear Information System (INIS)

    Zeldovich, Y.B.

    1983-01-01

    This paper fives a general review of modern cosmology. The following subjects are discussed: hot big bang and periodization of the evolution; Hubble expansion; the structure of the universe (pancake theory); baryon asymmetry; inflatory universe. (Auth.)

  11. Influence of texture on fracture toughness of zircaloy cladding

    International Nuclear Information System (INIS)

    Grigoriev, V.; Andersson, Stefan

    1997-06-01

    The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill's theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture

  12. Research on laser cladding control system based on fuzzy PID

    Science.gov (United States)

    Zhang, Chuanwei; Yu, Zhengyang

    2017-12-01

    Laser cladding technology has a high demand for control system, and the domestic laser cladding control system mostly uses the traditional PID control algorithm. Therefore, the laser cladding control system has a lot of room for improvement. This feature is suitable for laser cladding technology, Based on fuzzy PID three closed-loop control system, and compared with the conventional PID; At the same time, the laser cladding experiment and friction and wear experiment were carried out under the premise of ensuring the reasonable control system. Experiments show that compared with the conventional PID algorithm in fuzzy the PID algorithm under the surface of the cladding layer is more smooth, the surface roughness increases, and the wear resistance of the cladding layer is also enhanced.

  13. Comparative analysis of mechanical characteristics of solidified concentrates from BWR system using Yugoslav and Italian cements

    International Nuclear Information System (INIS)

    Plecas, I.; Peric, A.; Drljaca, J.; Kostadinovic, A.

    1987-01-01

    In this paper, properties of Italian and Yugoslav cement mixture with BWR evaporation concentrates were compared, research was held upon fifteen samples, according to the adequate formulations. Samples were made in standard cube form, side 10 cm. Functional relationship between decreasing the compressive strength and amount of incorporated BWR concentrate cement mixture was developed. The results of research showed nearly the same mechanical properties of solidified BWR concentrate with Italian and Yugoslav cements. (author)

  14. Kinematics of two-phase mixture level motion in BWR pressure vessels

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Mallen, A.N.; Stritar, A.

    1985-01-01

    A model is presented for predicting two-phase mixture level elevations in BWR systems. The model accounts for the particular geometry and conditions in a BWR system during Small-Break Loss of Coolant Accidents. The model presented here is particularly suitable for efficient, high-speed simulations on small minicomputers. The model has been implemented and tested. Results are shown from BWR ATWS simulations

  15. Demonstration of fuel resistant to pellet-cladding interaction: Phase 2. Fourth semiannual report, July-December 1980

    International Nuclear Information System (INIS)

    Rosenbaum, H.S.

    1981-03-01

    This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Two fuel concepts have been developed for possible demonstration: (a) Cu-barrier fuel and (b) Zr-liner fuel. These advanced fuels (known collectively as barrier fuels) have special fuel cladding designed to avoid the harmful effects of localized stress and reactive fission products during reactor service. Within the scope of this program one of these concepts had to be selected for a large-scale demonstration in a commercial power reactor. The selection was made to demonstrate Zr-liner fuel and to include bundles which have liners prepared from either low oxygen sponge zirconium or of crystal bar zirconium. The demonstration is intended to include a total of 132 barrier bundles in the reload for Quad Cities Unit 2, Cycle 6. In the current report period changes in the nuclear design were made to respond to changes in the Energy Utilization Plan for Quad Cities Unit 2. Bundle designs were completed, and were licensed for use in a BWR/3. The core specific licensing will be done as part of the reload license for Quad Cities Unit 2, Cycle 6

  16. Loop capabilities in Rez for water chemistry and corrosion control of cladding and in-core components

    International Nuclear Information System (INIS)

    Kysela, J.; Zmitko, M.; Srank, J.; Vsolak, R.

    1999-01-01

    Main characteristics of LVR-15 research reactor and its irradiation facilities are presented. For testing of cladding, internals and RPV materials specialised loop are used. There are now five high pressure loops modelling PWR, WWER or BWR water environment and chemistry. Loops can be connected with instrumented in-pile channels enable slow strain rate testing, 1CT or 2CT specimens loading and electrically heated rods exposition. Reactor dosimetry including neutronic parameters measurements and calculations and mock-up experiments are used. Water chemistry control involves gas (O 2 , H 2 ) dosing system, Orbisphere H 2 /O 2 measurement, electrochemical potential (ECP) measurements and specialised analytical chemistry laboratory. For cladding corrosion studies in-pile channels with four electrically heated rods with heat flux up to 100 W/cm 2 , void fraction 5 % at the outlet, inlet temperature 320 deg. C and flow velocity 3 m/s were development and tested. For corrosion layer investigation there is eddy current measurements and PIE techniques which use crud thickness measurement, chemical analyses of the crud, optical metallography, hydrogen analysis, SEM and TEM. (author)

  17. Construction techniques and management methods for BWR plants

    International Nuclear Information System (INIS)

    Shimizu, Yohji; Tateishi, Mizuo; Hayashi, Yoshishige

    1989-01-01

    Toshiba is constantly striving for safer and more efficient plant construction to realize high-quality BWR plants within a short construction period. To achieve these aims, Toshiba has developed and improved a large number of construction techniques and construction management methods. In the area of installation, various techniques have been applied such as the modularization of piping and equipment, shop installation of reactor internals, etc. Further, installation management has been upgraded by the use of pre-installation review programs, the development of installation control systems, etc. For commissioning, improvements in commissioning management have been achieved through the use of computer systems, and testing methods have also been upgraded by the development of computer systems for the recording and analysis of test data and the automatic adjustment of controllers in the main control system of the BWR. This paper outlines these construction techniques and management methods. (author)

  18. Specifications of the BWR simulator for HAMMLAB 2000

    International Nuclear Information System (INIS)

    Grini, Rolf-Einar; Miettinen, Jaakko; Nurmilaukas, Pekka; Raussi; Pekka; Saarni, Ray; Stokke; Egil; Soerensen, Aimar; Tiihonen, Olli

    1998-02-01

    The Boiling Water Reactor (BWR) simulator for HAMMLAB 2000 will be a model of the Swedish plant Forsmark-3. This report gives the specifications of the BWR simulator. The bulk of the report is a copy of the relevant addendum to the contract with the developer, and to the contract with the group of utilities and with ABB Atom. After a general overview, each plant system is described one after the other (using the reference plant system coding), and the simulation of each system is specified. Even the systems that shall not be simulated are included; in those cases the specification is: It is not required that ... is simulated. A list of malfunctions is given, as well as a list of validation transients. Finally the operator interface is specified. (author)

  19. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    International Nuclear Information System (INIS)

    Chiang, Ren-Tai; Williams, John B.; Folk, Ken S.

    2008-01-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  20. BWR Fuel Assemblies Physics Analysis Utilizing 3D MCNP Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Ren-Tai [University of Florida, Gainesville, Florida 32611 (United States); Williams, John B.; Folk, Ken S. [Southern Nuclear Company, Birmingham, Alabama 35242 (United States)

    2008-07-01

    MCNP is used to model a partially controlled BWR fresh fuel four assemblies (2x2) system for better understanding BWR fuel behavior and for benchmarking production codes. The impact of the GE14 plenum regions on axial power distribution is observed by comparing against the GE13 axial power distribution, in which the GE14 relative power is lower than the GE13 relative power at the 15. node and at the 16. node due to presence of the plenum regions in GE14 fuel in these two nodes. The segmented rod power distribution study indicates that the azimuthally dependent power distribution is very significant for the fuel rods next to the water gap in the uncontrolled portion. (authors)

  1. Initiation model for intergranular stress corrosion cracking in BWR pipes

    International Nuclear Information System (INIS)

    Hishida, Mamoru; Kawakubo, Takashi; Nakagawa, Yuji; Arii, Mitsuru.

    1981-01-01

    Discussions were made on the keys of intergranular stress corrosion cracking of austenitic stainless steel in high-temperature water in laboratories and stress corrosion cracking incidents in operating plants. Based on these discussions, a model was set up of intergranular stress corrosion cracking initiation in BWR pipes. Regarding the model, it was presumed that the intergranular stress corrosion cracking initiates during start up periods whenever heat-affected zones in welded pipes are highly sensitized and suffer dynamic strain in transient water containing dissolved oxygen. A series of BWR start up simulation tests were made by using a flowing autoclave system with slow strain rate test equipment. Validity of the model was confirmed through the test results. (author)

  2. A BWR 24-month cycle analysis using multicycle techniques

    International Nuclear Information System (INIS)

    Hartley, K.D.

    1993-01-01

    Boiling water reactor (BWR) fuel cycle design analyses have become increasingly challenging in the past several years. As utilities continue to seek improved capacity factors, reduced power generation costs, and reduced outage costs, longer cycle lengths and fuel design optimization become important considerations. Accurate multicycle analysis techniques are necessary to determine the viability of fuel designs and cycle operating strategies to meet reactor operating requirements, e.g., meet thermal and reactivity margin constraints, while minimizing overall fuel cycle costs. Siemens Power Corporation (SPC), Nuclear Division, has successfully employed multi-cycle analysis techniques with realistic rodded cycle depletions to demonstrate equilibrium fuel cycle performance in 24-month cycles. Analyses have been performed by a BWR/5 reactor, at both rated and uprated power conditions

  3. Standard Technical Specifications, General Electric plants, BWR/4

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for General Electric Plants, BWR/4, and documents the positions of the Nuclear Regulatory Commission based on the BWR Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power specifications. This report contains three volumes. This document, Volume 2, contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS

  4. Neutron noise analysis of BWR using time series analysis

    International Nuclear Information System (INIS)

    Fukunishi, Kohyu

    1976-01-01

    The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW) [de

  5. Condensate polishing guidelines for PWR and BWR plants

    International Nuclear Information System (INIS)

    Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

    1993-03-01

    Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

  6. Vertical Drop of 44-BWR Waste Package With Lifting Collars

    Energy Technology Data Exchange (ETDEWEB)

    A.K. Scheider

    2005-08-23

    The objective of this calculation is to determine the structural response of a waste package (WP) dropped flat on its bottom from a specified height. The WP used for that purpose is the 44-Boiling Water Reactor (BWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. The Uncanistered Waste Disposal Container System is classified as Quality Level 1 (Ref. 4, page 7). Therefore, this calculation is subject to the requirements of the Quality Assurance Requirements and Description (Ref. 16). AP-3. 12Q, Design Calculations and Analyses (Ref. 11) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 44-BWR WP considered in this calculation and provides the potential dimensions and materials for that design.

  7. Paired replacement fuel assemblies for BWR-type reactor

    International Nuclear Information System (INIS)

    Oguchi, Kazushige.

    1997-01-01

    There are disposed a large-diameter water rod constituting a non-boiling region at a central portion and paired replacement fuel assemblies for two streams having the same average enrichment degree and different amount of burnable poisons. The paired replacement fuel assemblies comprise a first fuel assembly having a less amount of burnable poisons and a second fuel assembly having a larger amount of burnable poisons. A number of burnable poison-containing fuel rods in adjacent with the large diameter water rod is increased in the second fuel assembly than the first fuel assembly. Then, the poison of the paired replacement fuel assemblies for the BWR type reactor can be annihilated simultaneously at the final stage of the cycle. Accordingly, fuels for a BWR type reactor excellent in economical property and safety and facilitating the design of the replacement reactor core can be obtained. (N.H.)

  8. Review of international solutions to NEACRP benchmark BWR lattice cell problems

    International Nuclear Information System (INIS)

    Halsall, M.J.

    1977-12-01

    This paper summarises international solutions to a set of BWR benchmark problems. The problems, posed as an activity sponsored by the Nuclear Energy Agency Committee on Reactor Physics, were as follows: 9-pin supercell with central burnable poison pin, mini-BWR with 4 pin-cells and water gaps and control rod cruciform, full 7 x 7 pin BWR lattice cell with differential U 235 enrichment, and full 8 x 8 pin BWR lattice cell with water-hole, Pu-loading, burnable poison, and homogenised cruciform control rod. Solutions have been contributed by Denmark, Japan, Sweden, Switzerland and the UK. (author)

  9. Evaluation of PWR and BWR pin cell benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Pilgroms, B.J.; Gruppelaar, H.; Janssen, A.J. (Netherlands Energy Research Foundation (ECN), Petten (Netherlands)); Hoogenboom, J.E.; Leege, P.F.A. de (Interuniversitair Reactor Inst., Delft (Netherlands)); Voet, J. van der (Gemeenschappelijke Kernenergiecentrale Nederland NV, Dodewaard (Netherlands)); Verhagen, F.C.M. (Keuring van Electrotechnische Materialen NV, Arnhem (Netherlands))

    1991-12-01

    Benchmark results of the Dutch PINK working group on the PWR and BWR pin cell calculational benchmark as defined by EPRI are presented and evaluated. The observed discrepancies are problem dependent: a part of the results is satisfactory, some other results require further analysis. A brief overview is given of the different code packages used in this analysis. (author). 14 refs.; 9 figs.; 30 tabs.

  10. Latest experiences in inspecting the inside of BWR vessel shields

    Energy Technology Data Exchange (ETDEWEB)

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  11. Results of modeling advanced BWR fuel designs using CASMO-4

    International Nuclear Information System (INIS)

    Knott, D.; Edenius, M.

    1996-01-01

    Advanced BWR fuel designs from General Electric, Siemens and ABB-Atom have been analyzed using CASMO-4 and compared against fission rate distributions and control rod worths from MCNP. Included in the analysis were fuel storage rack configurations and proposed mixed oxide (MOX) designs. Results are also presented from several cycles of SIMULATE-3 core follow analysis, using nodal data generated by CASMO-4, for cycles in transition from 8x8 designs to advanced fuel designs. (author)

  12. Thermohydraulic stability coupled to the neutronic in a BWR

    International Nuclear Information System (INIS)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A.; Castlllo D, R.

    2006-01-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the Laguna Verde

  13. TVA experience in BWR reload design and licensing

    International Nuclear Information System (INIS)

    Robertson, J.D.

    1986-01-01

    TVA has developed and implemented the capability to perform BWR reload core design and licensing analyses. The advantages accruing from this capability include the tangible cost-savings from performing reload analyses in-house. Also, ''intangible'' benefits such as increased operating flexibility and the ability to accommodate multivendor fuel designs have been demonstrated. The major disadvantage with performing in-house analyses is the cost associated with development and maintenance of the analytical methods and staff expertise

  14. Maximum thermal loading test of BWR fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Yoshitaka; Yoshimura, Kunihiro; Nakamura, Satoshi; Ishizuka, Takao.

    1987-01-01

    Various proving tests on the reliability of nuclear power plants have been conducted at the Nuclear Power Engineering Test Center and at the Japan Power Plant Engineering and Inspection Corporation. The tests were initiated at the request of the Ministry of International Trade and Industry (MITI). Toshiba undertook one of the proving tests on the reliability of nuclear fuel assembly; the maximum thermal loading test of BWR fuel assembly from the Nuclear Power Engineering Test Center. These tests are part of the proving tests mentioned above, and their purpose is to confirm the reliability of the thermal hydraulic engineering techniques. Toshiba has been engaged for the past nine years in the design, fabrication and testing of the equipment. For the project, a test model fuel assembly was used to measure the critical power of the BWR fuel assembly and the void and fluidity of the coolant. From the test results, it has been confirmed that the heat is transferred safely from the fuel assembly to the coolant in the BWR nuclear power plant. In addition, the propriety and reliability of the thermal hydraulic engineering techniques for the fuel assembly have been proved. (author)

  15. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  16. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  17. Seismic proving test of BWR primary loop recirculation system

    International Nuclear Information System (INIS)

    Sato, H.; Shigeta, M.; Karasawa, Y.

    1987-01-01

    The seismic proving test of BWR Primary Loop Recirculation system is the second test to use the large-scale, high-performance vibration table of Tadotsu Engineering Laboratory. The purpose of this test is to prove the seismic reliability of the primary loop recirculation system (PLR), one of the most important safety components in the BWR nuclear plants, and also to confirm the adequacy of seismic analysis method used in the current seismic design. To achieve the purpose, the test was conducted under conditions and scale as near as possible to actual systems. The strength proving test was carried out with the test model mounted on the vibration table in consideration of basic design earthquake ground motions and other conditions to confirm the soundness of structure and the strength against earthquakes. Detailed analysis and analytic evaluation of the data obtained from the test was conducted to confirm the adequacy of the seismic analysis method and earthquake response analysis method used in the current seismic design. Then, on the basis of the results obtained, the seismic safety and reliability of BWR primary loop recirculation of the actual plants was fully evaluated

  18. Novel modular natural circulation BWR design and safety evaluation

    International Nuclear Information System (INIS)

    Ishii, Mamoru; Shi, Shanbin; Yang, Won Sik; Wu, Zeyun; Rassame, Somboon; Liu, Yang

    2015-01-01

    Highlights: • Introduction of BWR-type natural circulation small modular reactor preliminary design (NMR-50). • Design of long fuel cycle length for the NMR-50. • Design of double passive safety systems for the NMR-50. • RELAP5 analyses of design basis accidents for the NMR-50. - Abstract: The Purdue NMR (Novel Modular Reactor) represents a BWR-type small modular reactor with a significantly reduced reactor pressure vessel (RPV) height. Specifically, it has one third the height of a conventional BWR RPV with an electrical output of 50 MWe. The preliminary design of the NMR-50 including reactor, fuel cycle, and safety systems is described and discussed. The improved neutronics design of the NMR-50 extends the fuel cycle length up to 10 years. The NMR-50 is designed with double passive engineering safety system, which is intended to withstand a prolonged station black out with loss of ultimate heat sink accident such as experienced at Fukushima. In order to evaluate the safety features of the NMR-50, two representative design basis accidents, i.e. main steam line break (MSLB) and bottom drain line break (BDLB), are simulated by using the best-estimate thermal–hydraulic code RELAP5. The RPV water inventory, containment pressure, and the performance of engineering safety systems are investigated for about 33 h after the initiation of the accidents

  19. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  20. Sensitivity of BWR shutdown margin tests to local reactivity anomalies

    International Nuclear Information System (INIS)

    Cokinos, D.M.; Carew, J.F.

    1987-01-01

    Successful shutdown margin (SDM) demonstration is a required procedure in the startup of a newly configured boiling water reactor (BWR) core. In its most reactive condition throughout a cycle, a BWR core must be capable of being made subcritical by a specified margin with the highest worth control rod fully withdrawn and all other rods at their fully inserted positions. Two different methods are used to demonstrate SDM: (a) the adjacent-rod test and (b) the in-sequence test. In the adjacent-rod test, the strongest rod is fully withdrawn and an adjacent rod is withdrawn to reach criticality. In the in-sequence test, control rods spread throughout the core are withdrawn in a predetermined sequence of withdrawals. Larger than expected core k/sub eff/ values have been observed during the performance of BWR SDM tests. The purpose of the work summarized in this paper has been to investigated and quantify the sensitivity of both the adjacent-rod and in-sequence SDM tests to local reactivity anomalies. This was accomplished by introducing reactivity perturbations at selected four-bundle cell locations and by evaluating their effect on core reactivity in each of the two tests

  1. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    International Nuclear Information System (INIS)

    Morikawa, Yoshitake

    1995-01-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data

  2. Reduction of radiation exposure in Japanese BWR Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Morikawa, Yoshitake [ISOGO Nuclear Engineering Center, Yokohama (Japan)

    1995-03-01

    The reduction of occupational exposure to radiation during the annual inspection and maintenance outages of Japanese boiling water reactors (BWR) is one of the most important objectives for stable and reliable operation. It was shown that this radiation exposure is caused by radionuclides, such as Co-60, Co-58 and Mn-54 which are produced from the metal elements Co, Ni, and Fe present in the corrosion products of structural materials that had been irradiated by neutrons. Therefore, to reduce radiation sources and exposures in Japanese BWRs, attempts have been reinforced to remove corrosion products and activated corrosion products from the primary coolant system. This paper describes the progress of the application of these measures to Japanese BWRs. Most Japanese BWR-4 and BWR-5 type nuclear power plants started their commercial operations during the 1970s. With the elapse of time during operations, a problem came to the forefront, namely that occupational radiation exposure during plant outages gradually increased, which obstructed the smooth running of inspections and maintenance work. To overcome this problem, extensive studies to derive effective countermeasures for radiation exposure reduction were undertaken, based on the evaluation of the plants operation data.

  3. Advanced chemistry management system to optimize BWR chemistry control

    International Nuclear Information System (INIS)

    Maeda, K.; Nagasawa, K.

    2002-01-01

    BWR plant chemistry control has close relationships among nuclear safety, component reliability, radiation field management and fuel integrity. Advanced technology is required to improve chemistry control [1,3,6,7,10,11]. Toshiba has developed TACMAN (Toshiba Advanced Chemistry Management system) to support BWR chemistry control. The TACMAN has been developed as response to utilities' years of requirements to keep plant operation safety, reliability and cost benefit. The advanced technology built into the TACMAN allows utilities to make efficient chemistry control and to keep cost benefit. TACMAN is currently being used in response to the needs for tools those plant chemists and engineers could use to optimize and identify plant chemistry conditions continuously. If an incipient condition or anomaly is detected at early stage, root causes evaluation and immediate countermeasures can be provided. Especially, the expert system brings numerous and competitive advantages not only to improve plant chemistry reliability but also to standardize and systematize know-how, empirical knowledge and technologies in BWR chemistry This paper shows detail functions of TACMAN and practical results to evaluate actual plant. (authors)

  4. Large bundle BWR test CORA-18: Test results

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

    1998-04-01

    The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

  5. Parametric study of the behaviour of a pre irradiated BWR fuel rod under conditions of LOCA simulated in the halden in pile test system with the FALCON code

    Energy Technology Data Exchange (ETDEWEB)

    Khvostov, G.; Zimmermann, M. A. [Laboratory for Reactor Physics and Systems Behaviour, Paul Scherrer Institut, Villigen (Switzerland); Ledergerber, G. [Kernkraftwerk Leibstadt AG, Leibstadt (Switzerland); Kolstad, E. [Institute for Energy Technology - OECD Halden Reactor Project, Halden (Norway); Montgomery, R. O. [Anatech Corporation, San Diego (United States)

    2008-10-15

    A new LOCA test at Halden was planned as the first experiment within the Halden LOCA program addressing the behaviour of commercially irradiated BWR fuel of medium burn up with burst of the cladding expected to occur at a temperature of about 1050.deg.C, which is essentially higher than in the preceding experiments. The specific measures to be adopted have been suggested based upon a parametric study using the FALCON fuel behaviour code and aimed at an optimized design of the test fuel rod for the given high target cladding temperature of 1150 .deg. C (peak local). The analysis has shown a reasonable agreement with the fundamental experimental findings, such as correlations of NUREG 0630, as well as consistency with the data from Halden LOCA testing available so far. Thus, a general conclusion is drawn about the applicability of the methodology developed at PSI to the analysis of LWR fuel rod behaviour during LOCA, in consideration of the effects of fuel burn up.

  6. Phased array concept for the ultrasonic inservice inspection of the spherical bottom of BWR-pressure vessels

    International Nuclear Information System (INIS)

    Brekow, G.; Wuestenberg, H.; Moehrle, W.; Schulz, E.

    1989-01-01

    The spherical bottom of BWR-pressure vessels contains holes for the nozzles of control rods and instrumentation. Up to now the detectable areas for the ultrasonic inspection are the accessible ligaments between the nozzles with an orientation parallel and transverse to the manipulator rails. Some licensing authorities demand an inspection technique capable of reliably detecting significant crack initiation in all critical areas near the cladding of the spherical inner surface. By order and in cooperation with the HEW we have developed a computer controlled equipment with two ultrasonic probes containing four linear arrays and a digitized A-scan storage for documentation and evaluation of inspection results. The manipulator guided probe movement in the paths between the nozzles of the spherical bottom is controlled by a computer program. This program determines for each array system and for each coupling position the beam angle as a function of the variable skewing angle to realize detection conditions suited to possible crack positions at the longitudinal, transverse and diagonal ligaments between the nozzles for control rods and instrumentation. (orig./HP)

  7. Cross-sections for homogenized BWR fuel elements in 2d-diffusion theory by 1d-transport calculations

    International Nuclear Information System (INIS)

    Ambrosius, G.

    1980-01-01

    Leakage has a large influence on the thermal spectrum in a fuel rod cell of a BWR and originates: a) from rods with different absorptions and; b) from the different distances to the water gaps. Due to reason a) Gd-rods are treated together with a ring of the homogenized eight nearest neighbours. The often used definition of homogenized cross-sections as the ratio of the integrated reaction rate to the integrated flux proved to be inadequate. This homogenization method is exact as far as the flux is constant over the boundary and as the leakag e during calculating the homogenized cross-sections is similar to that during application. With respect to the condition b) a 1d-transport calculation for the whole fuel element with rings or slabs of homogenized fuel rod cells is performed. With the definition above the flux distribution is that of the fluxes in the moderator regions. The spectrum within each fuel rod cell which includes the leakage is calculated by superimposing at each energy on the flux distribution in the cell the flux at the cell position from the bundle calculation. Changes in the flux ratio between fuel and moderator due to the leakage are taken into account in a final few group 2d-diffusion calculation with fuel and (moderator + cladding) taken separately

  8. Thermohydraulic stability coupled to the neutronic in a BWR; Estabilidad termohidraulica acoplada a la neutronica en un BWR

    Energy Technology Data Exchange (ETDEWEB)

    Calleros M, G.; Zapata Y, M.; Gomez H, R.A.; Mendez M, A. [Comision Federal de Electricidad, Central Nucleoelectrica de Laguna Verde, Carretera Cardel-Nautla Km. 42.5, Mpio. Alto Lucero, Veracruz (Mexico); Castlllo D, R. [ININ, Carretera Mexico-Toluca Km 36.5, La Marquesa, Estado de Mexico (Mexico)]. e-mail: gcm9acpp@cfe.gob.mx

    2006-07-01

    In a BWR type reactor the phenomenon of the nuclear fission is presented, in which are liberated in stochastic form neutrons, originating that the population of the same ones varies in statistic form around a mean value. This variation will cause that when the neutron flow impacts on the neutron detectors, its are had as a result neutron flow signals with fluctuations around an average value. In this article it is shown that it conforms it lapses the time, this variations in the neutron flow (and therefore, in the flow signal due only to the fission), they presented oscillations inside a stable range, which won't be divergent. Considering that the BWR is characterized because boiling phenomena are presented, which affect the moderation of the neutrons, additional variations will be had in the signal coming from the neutron detectors, with relationship to the fission itself, which will be influenced by the feedback of the moderator's reactivity and of the temperature of the fuel pellet. Also, as the BWR it has coupled control systems to maintain the coolant level one and of the thermal power of the reactor, for each control action it was affected the neutron population. This means that the reactor could end up straying of a stable state condition. By it previously described, the study of the thermohydraulic stability coupled to the neutronic is complex. In this work it is shown the phenomenology, the mathematical models and the theoretical behavior associated to the stability of the BWR type reactor; the variables that affect it are identified, the models that reproduce the behavior of the thermohydraulic stability coupled to the neutronic, the way to maintain stable the reactor and the instrumentation that can settle to detect and to suppress uncertainties is described. In particular, is make reference to the evolution of the methods to maintain the stability of the reactor and the detection system and suppression of uncertainties implemented in the

  9. Demonstration of fuel resistant to pellet-cladding interaction. First semiannual report, July-December 1977

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, H.S. (comp.)

    1978-02-01

    Objective is the demonstration od advanced fuel concepts that are resistant to the failure mechanism known as fuel pellet-cladding interaction (PCI). Since currently used fuel in the nuclear power industry is subject to the PCI failure mechanism, reactor operators limit the rates of power increases and thus reduce their capacity factors in order to protect the fuel. Two barrier concepts are being prepared for demonstration: (a) Cu-Barrier fuel and (b) Zr-Liner fuel. The large-scale demonstration of the PCI-resistant fuel is being designed generically to show feasibility of such a demonstration in a commercial power reactor of type BWR/3 having a steady-state core. Using the core of Quad Cities-1 reactor at the beginning of Cycle 6, the insertion of the demonstration PCI-resistant fuel and the reactor operational plan are being designed. Support laboratory tests to date for the Demonstration have shown that these barrier fuels (both the Cu-Barrier and the Zr-Liner types) are resistant to PCI. Four lead test assemblies (LTA) of the advanced PCI-resistant fuel are being fabricated for insertion into the Quad Cities-1 Boiling Water Reactor at the beginning of Cycle 5 (January 1979).

  10. Spent fuel cladding containment credit test

    International Nuclear Information System (INIS)

    Wilson, C.N.

    1983-01-01

    As an initial step in addressing the effectiveness of breached cladding as a barrier to radionuclide release from the repository during the post-containment period, preliminary scoping tests have been initiated which compare radionuclide releases from spent fuel specimens with artificially induced cladding defects of various severities. The artificially induced defects are all more severe than the typical in-reactor type breaches which are expected to be the principal type of breach entering the repository for terminal storage. These preliminary scoping tests being conducted by Westinghouse Hanford Company for the Lawrence Livermore National Laboratory Waste Package Development Program in support of the Tuff repository project at the Nevada Test Site are described. Also included in this presentation are selected initial results from these tests. 22 figures

  11. Automatic welding and cladding in heavy fabrication

    International Nuclear Information System (INIS)

    Altamer, A. de

    1980-01-01

    A description is given of the automatic welding processes used by an Italian fabricator of pressure vessels for petrochemical and nuclear plant. The automatic submerged arc welding, submerged arc strip cladding, pulsed TIG, hot wire TIG and MIG welding processes have proved satisfactory in terms of process reliability, metal deposition rate, and cost effectiveness for low alloy and carbon steels. An example shows sequences required during automatic butt welding, including heat treatments. Factors which govern satisfactory automatic welding include automatic anti-drift rotator device, electrode guidance and bead programming system, the capability of single and dual head operation, flux recovery and slag removal systems, operator environment and controls, maintaining continuity of welding and automatic reverse side grinding. Automatic welding is used for: joining vessel sections; joining tubes to tubeplate; cladding of vessel rings and tubes, dished ends and extruded nozzles; nozzle to shell and butt welds, including narrow gap welding. (author)

  12. Creep collapse of TAPS fuel cladding

    International Nuclear Information System (INIS)

    Chaudhry, S.M.; Anand, A.K.

    1975-01-01

    Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported (internally) at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : (1) inelastic collapse of perfectly circular tubes (2) plastic instability in oval tubes (3) effect of creep on ovality. Creep is considered to be a non-linear combination of the following : (a) thermal creep (b) intresenic creep (c) stress aided radiation enhanced (d) stress free growth (4) Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. (author)

  13. Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction (FCCI)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction (Fuel Cladding Chemical Interaction, FCCI) with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.

  14. LASER CLADDING ON ALUMINIUM BASE ALLOYS

    OpenAIRE

    Pilloz , M.; Pelletier , J.; Vannes , A.; Bignonnet , A.

    1991-01-01

    laser cladding is often performed on iron or titanium base alloys. In the present work, this method is employed on aluminum alloys ; nickel or silicon are added by powder injection. Addition of silicon leads to sound surface layers, but with moderated properties, while the presence of nickel induces the formation of hard intermetallic compounds and then to an attractive hardening phenomena ; however a recovery treatment has to be carried out, in order to eliminate porosity in the near surface...

  15. Nuclear-powered pacemaker fuel cladding study

    International Nuclear Information System (INIS)

    Shoup, R.L.

    1976-01-01

    The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice

  16. Degradation resistant fuel cladding materials and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC (United States); Montes, J. [ENUSA, Madrid (Spain)

    1995-12-31

    GE has been producing the degradation resistant cladding (zirconium liner and zircaloy-2 surface larger) described here with the cooperation of its primary zirconium vendors since the beginning of 1994. Approximately 24 fuel reloads, or in excess of 250,000 fuel rods, have been produced using this material by GE. GE has also produced tubing for one reload of fuel that is currently being produced by its technology affiliate ENUSA. (orig./HP)

  17. Corrosion behaviour of laser clad stainless steels

    International Nuclear Information System (INIS)

    Damborenea, J.J. de; Weerasinghe, V.M.; West, D.R.F.

    1993-01-01

    The present paper is focussed in the study of the properties of a clad layer of stainless steel on a mild steel. By blowing powder of the alloy into a melt pool generated by a laser of 2 KW, an homogeneous layer of 316 stainless steel can be obtained. Structure, composition and corrosion behaviour are similar to those of a stainless steel in as-received condition. (Author)

  18. Plasma spheroidization and cladding of powders

    Energy Technology Data Exchange (ETDEWEB)

    Petrunichev, V.A.; Averin, V.V.; Sorokin, L.M.; Koroleva, E.B.

    1987-02-01

    With reference to experimental results for nickel and chromium alloys, it is shown that complex alloy powders can be spheroidized in plasma discharges using an argon plasma with hydrogen. The spheroidizing process is accompanied by the reduction of surface oxides, with uniform element distribution within the particles; the granulometric composition of the particles is preserved. It is also shown that plasma technology can be used for producing metal-clad oxide and carbide powders, which improve the performance of cermets and coatings.

  19. Alloy development for high burnup cladding (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    An overview on current alloy development for high burnup PWR fuel cladding is given. It is mainly based on literature data. First, the reasons for an increase of the current mean discharge burnup from 35 MWd / kg(U) to 70 MWd / kg(U) are outlined. From the material data, it is shown that a batch average burnup of 60-70 MWd / kg(U), as aimed by many fuel vendors, can not be achieved with stand (=ASTM-) Zry-4 cladding tubes without violating accepted design criteria. Specifically criteria which limit maximum oxide scale thickness and maximum hydrogen content, and to a less degree, maximum creep and growth rate, can not be achieved. The development potential of standard Zry-4 is shown. Even when taking advantage of this potential, it is shown that an 'improved' Zry-4 is reaching its limits when it achieves the target burnup. The behavior of some Zr alloys outside the ASTM range is shown, and the advantages and disadvantages of the 3 alloy groups (ZrSn+transition metals, ZrNb, ZrSnNb+transition metals) which are currently considered to have the development potential for high burnup cladding materials are depicted. Finally, conclusions are drawn. (author). 14 refs., 11 tabs., 82 figs.

  20. Advanced LWR Nuclear Fuel Cladding Development

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Griffith, G.

    2012-01-01

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

  1. Cladding failure by local plastic instability

    International Nuclear Information System (INIS)

    Kramer, J.M.; Deitrich, L.W.

    1977-01-01

    Cladding failure is one of the major considerations in analysis of fast-reactor fuel pin behavior during hypothetical accident transients since time, location and nature of failure govern the early post-failure material motion and reactivity feedback. Out-of-Pile transient burst tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture. To investigate the details of cladding bulging, a perturbation analysis of the equations governing the large deformation of a cylindrical shell has been developed. The overall deformation history is assumed to consist of a small perturbation epsilon of the radial displacement superimposed on large axisymmetric displacements. Computations have been carried out using high temperature properties of stainless steel in conjunction with various constitutive theories, including a generalization of the Endochronic Theory of Plasticity in which both time-independent and time-dependent plastic strains are modeled. Although the results of the calculations are all qualitatively similar, it appears that modeling of both time-independent and time-dependent plastic strains is necessary to interpret the transient burst test results. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or non-symmetric cooling. (Auth.)

  2. Plant analyzer for high-speed interactive simulation of BWR plant transients

    International Nuclear Information System (INIS)

    Cheng, H.S.; Lekach, S.V.; Mallen, A.N.; Wulff, W.; Cerbone, R.J.

    1984-01-01

    A combination of advanced modeling techniques and modern, special-purpose peripheral minicomputer technology was utilized to develop a plant analyzer which affords realistic predictions of plant transients and severe off-normal events in LWR power plants through on-line simulations at speeds up to 10 times faster than actual process speeds. The mathematical models account for nonequilibrium, nonhomogeneous two-phase flow effects in the coolant, for acoustical effects in the steam line and for the dynamics of the entire balance of the plant. Reactor core models include point kinetics with reactivity feedback due to void fraction, fuel temperature, coolant temperature, and boron concentration as well as a conduction model for predicting fuel and clad temperatures. Control systems and trip logic for plant protection systems are also simulated. The AD10 of Applied Dynamics International, a special-purpose peripheral processor, is used as the principal hardware of the plant analyzer

  3. Italian Modernities

    DEFF Research Database (Denmark)

    Thomassen, Bjørn; Forlenza, Rosario

    assumptions that have substituted for thought and that have perpetuated prejudices both within and outside Italy’s borders. Grounded in meticulous historical and ethnological research, Italian Modernities deserves as wide an audience as its scholarship is deep.” (Michael Herzfeld, Ernest E. Monrad Professor...

  4. Montreal Modern

    DEFF Research Database (Denmark)

    Handberg, Kristian

    2015-01-01

    , and the space age modernism of the 1960s following the Expo 67 and Quebec’s Quiet Revolution. This is reflected in the city’s thriving retro culture through the study of two groups of retro shops. In circulating specific memories and objects in a specific context, retro is an important negotiation of the past...

  5. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    Salaices, M.; Salaices, E.; Ovando, R.; Esquivias, J.

    2011-11-01

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  6. Microstructure and Mechanical Properties of Laser Clad and Post-cladding Tempered AISI H13 Tool Steel

    Science.gov (United States)

    Telasang, Gururaj; Dutta Majumdar, Jyotsna; Wasekar, Nitin; Padmanabham, G.; Manna, Indranil

    2015-05-01

    This study reports a detailed investigation of the microstructure and mechanical properties (wear resistance and tensile strength) of hardened and tempered AISI H13 tool steel substrate following laser cladding with AISI H13 tool steel powder in as-clad and after post-cladding conventional bulk isothermal tempering [at 823 K (550 °C) for 2 hours] heat treatment. Laser cladding was carried out on AISI H13 tool steel substrate using a 6 kW continuous wave diode laser coupled with fiber delivering an energy density of 133 J/mm2 and equipped with a co-axial powder feeding nozzle capable of feeding powder at the rate of 13.3 × 10-3 g/mm2. Laser clad zone comprises martensite, retained austenite, and carbides, and measures an average hardness of 600 to 650 VHN. Subsequent isothermal tempering converted the microstructure into one with tempered martensite and uniform dispersion of carbides with a hardness of 550 to 650 VHN. Interestingly, laser cladding introduced residual compressive stress of 670 ± 15 MPa, which reduces to 580 ± 20 MPa following isothermal tempering. Micro-tensile testing with specimens machined from the clad zone across or transverse to cladding direction showed high strength but failure in brittle mode. On the other hand, similar testing with samples sectioned from the clad zone parallel or longitudinal to the direction of laser cladding prior to and after post-cladding tempering recorded lower strength but ductile failure with 4.7 and 8 pct elongation, respectively. Wear resistance of the laser surface clad and post-cladding tempered samples (evaluated by fretting wear testing) registered superior performance as compared to that of conventional hardened and tempered AISI H13 tool steel.

  7. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    International Nuclear Information System (INIS)

    Garcia, S.E.; Giannelli, J.F.; Jarvis, M.L.

    2010-01-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  8. BWR chemistry control status: a summary of industry chemistry status relative to the BWR water chemistry guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, M.L., E-mail: jgiannelli@finetech.com [Finetech, Inc., Parsippany, NJ (United States)

    2010-07-01

    The EPRI Boiling Water Reactor (BWR) Water Chemistry Guidelines were revised and issued in October 2008. The 2008 Revision of the Guidelines continues to focus on intergranular stress corrosion cracking (IGSCC), which can limit the service life of susceptible materials and components exposed to water chemistry environments. The 2008 Revision also places increased emphasis on fuel performance and meeting the industry goal of zero fuel failures by 2010. As an industry consensus document, the Guidelines were created to provide proactive water chemistry control strategies for mitigating IGSCC, maintaining fuel integrity and controlling radiation fields. The Guidelines provide a technically-based framework for an effective BWR water chemistry program. This paper provides an overview of industry experience relative to the Guidelines. Over the past few years, many BWR units have implemented noble metal chemical application technologies either during plant hot or cold shutdown or at normal power operating conditions. This paper explores plant experience with optimized water chemistry, implementation of various additive chemistries such as noble metal application and zinc addition, and compliance with the Guidelines recommendations. Depleted zinc oxide addition has been broadly applied across the BWR fleet since the 1980s. The guidance for zinc addition has been revised in the Guidelines to reflect concerns with fuel performance. While zinc addition is a successful method for shutdown dose rate control, concerns still exist for high zinc deposition on fuel surfaces, especially when feedwater iron is elevated and as fuel cores are being driven to provide maximum power output over longer fuel cycles. Recent plant experience has shown that the utilization of online noble metal application and continued zinc addition may provide additional benefits for radiation control. Dose rate experiences at plants utilizing the online noble metal application technology and zinc addition

  9. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  10. Impact of reactor water chemistry on cladding performance

    International Nuclear Information System (INIS)

    Cox, B.

    1997-01-01

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

  11. Impact of reactor water chemistry on cladding performance

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  12. Research Progress on Laser Cladding Amorphous Coatings on Metallic Substrates

    Directory of Open Access Journals (Sweden)

    CHEN Ming-hui

    2017-01-01

    Full Text Available The microstructure and property of amorphous alloy as well as the limitations of the traditional manufacturing methods for the bulk amorphous alloy were briefly introduced in this paper.Combined with characteristics of the laser cladding technique,the research status of the laser cladding Fe-based,Zr-based,Ni-based,Cu-based and Al-based amorphous coatings on the metal substrates were mainly summarized.The effects of factors such as laser processing parameter,micro-alloying element type and content and reinforcing phase on the laser cladding amorphous coatings were also involved.Finally,the main problems and the future research directions of the composition design and control of the laser-cladded amorphous coating,the design and optimization of the laser cladding process,and the basic theory of the laser cladding amorphous coatings were also put forward finally.

  13. Hygrothermal performance of ventilated wooden cladding

    Energy Technology Data Exchange (ETDEWEB)

    Nore, Kristine

    2009-10-15

    This project contributes to more accurate design guidelines for high-performance building envelopes by analysis of hygrothermal performance of ventilated wooden cladding. Hygrothermal performance is defined by cladding temperature and moisture conditions, and subsequently by risk of degradation. Wood cladding is the most common facade material used in rural and residential areas in Norway. Historically, wooden cladding design varied in different regions in Norway. This was due to both climatic variations and the logistical distance to materials and craftspeople. The rebuilding of Norwegian houses in the 1950s followed central guidelines where local climate adaptation was often not evaluated. Nowadays we find some technical solutions that do not withstand all climate exposures. The demand for thermal comfort and also energy savings has changed hygrothermal condition of the building envelopes. In well-insulated wall assemblies, the cladding temperature is lower compared to traditional walls. Thus the drying out potential is smaller, and the risk of decay may be higher. The climate change scenario indicates a warmer and wetter future in Norway. Future buildings should be designed to endure harsher climate exposure than at present. Is there a need for refined climate differentiated design guidelines for building enclosures? As part of the Norwegian research programme 'Climate 2000', varieties of wooden claddings have been investigated on a test house in Trondheim. The aim of this investigation was to increase our understanding of the relation between microclimatic conditions and the responding hygrothermal performance of wooden cladding, according to orientation, design of ventilation gap, wood material quality and surface treatment. The two test facades, facing east and west have different climate exposure. Hourly measurements of in total 250 sensors provide meteorological data; temperature, radiation, wind properties, relative humidity, and test house data

  14. The use of the partial coherence function technique for the investigation of BWR noise dynamics

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1983-01-01

    The extensive experimental investigations, at the last time, indicate that the partial coherence function technique can be a powerful method of the investigation of BWR noise dynamics. Symple BWR noise dynamics model for the global noise study, based on different noise phenomena, is proposed in this paper. (author)

  15. Analyses on Silicide Coating for LOCA Resistant Cladding

    Energy Technology Data Exchange (ETDEWEB)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings.

  16. Analyses on Silicide Coating for LOCA Resistant Cladding

    International Nuclear Information System (INIS)

    Sweidan, Faris B.; Lee, You Ho; Ryu, Ho Jin

    2015-01-01

    A particular focus of accident-tolerant fuel has been cladding due to the rapid high-temperature oxidation of zirconium-based cladding with the evolution of H2 when steam is a reactant. Some key features of the coated cladding include high-temperature resistance to oxidation, lower processing temperatures, and a high melting point of the coating. Zirconium alloys exhibit a reasonably high melting temperature, so a coating for the cladding is appealing if the coating increases the high-temperature resistance to oxidation. In this case, the cladding is protected from complete oxidation. The cladding coating involves the application of zirconium silicide onto Zr-based cladding. Zirconium silicide coating is expected to produce a glassy layer that becomes more protective at elevated temperature. For this reason, silicide coatings on cladding offer the potential for improved reliability at normal operating temperatures and at the higher transient temperatures encountered during accidents. Although ceramic coatings are brittle and may have weak points to be used as coating materials, several ceramic coatings were successful and showed adherent behavior and high resistance to oxidation. In this study, the oxidation behavior of zirconium silicide and its oxidation kinetics are analyzed. Zirconium silicide is a new suggested material to be used as coatings on existing Zr-based cladding alloys, the aim of this study is to evaluate if zirconium silicide is applicable to be used, so they can be more rapidly developed using existing cladding technology with some modifications. These silicide coatings are an attractive alternative to the use of coatings on zirconium claddings or to the lengthy development of monolithic ceramic or ceramic composite claddings and coatings

  17. Clad fiber capacitor and method of making same

    Science.gov (United States)

    Tuncer, Enis

    2012-12-11

    A clad capacitor and method of manufacture includes assembling a preform comprising a ductile, electrically conductive fiber; a ductile, electrically insulating cladding positioned on the fiber; and a ductile, electrically conductive sleeve positioned over the cladding. One or more preforms are then bundled, heated and drawn along a longitudinal axis to decrease the diameter of the ductile components of the preform and fuse the preform into a unitized strand.

  18. Pellet-clad interaction in water reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  19. Pellet-clad interaction in water reactor fuels

    International Nuclear Information System (INIS)

    2004-01-01

    The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. (A.L.B.)

  20. Computer analysis of elongation of the WWER fuel rod claddings

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2008-01-01

    In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. (authors)

  1. Laser cladding of Zr on Mg for improved corrosion properties

    International Nuclear Information System (INIS)

    Subramanian, R.; Sircar, S.; Mazumder, J.

    1989-01-01

    This paper reports the results of laser cladding of Mg-2wt%Zr, and Mg-5wt%Zr powder mixture onto magnesium. The microstructure of the laser clad was studied. From the microstructural study, the epitaxial regrowth of the clad region on the underlying substrate was observed. Martensite plates of different size were observed in transmission electron microscope for MG-2wt%Zr and Mg-5wt%Zr laser clad. The corrosion properties of the laser clad were evaluated in sea water (3.5% NaCl). The position of the laser claddings in the galvanic series of metals in sea water, the anodic polarization characteristics of the laser claddings and the protective nature and the stability of the passivating film formed have been determined. The formation of pits on the surface of the laser clad subjected to corrosion is reported. The corrosion properties of the laser claddings are compared with that of the commercially used magnesium alloy AZ91B

  2. Rectangular-cladding silicon slot waveguide with improved nonlinear performance

    Science.gov (United States)

    Huang, Zengzhi; Huang, Qingzhong; Wang, Yi; Xia, Jinsong

    2018-04-01

    Silicon slot waveguides have great potential in hybrid silicon integration to realize nonlinear optical applications. We propose a rectangular-cladding hybrid silicon slot waveguide. Simulation result shows that, with a rectangular-cladding, the slot waveguide can be formed by narrower silicon strips, so the two-photon absorption (TPA) loss in silicon is decreased. When the cladding material is a nonlinear polymer, the calculated TPA figure of merit (FOMTPA) is 4.4, close to the value of bulk nonlinear polymer of 5.0. This value confirms the good nonlinear performance of rectangular-cladding silicon slot waveguides.

  3. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  4. Determination of plastic anisotropy of zirconium alloys cladding

    International Nuclear Information System (INIS)

    Yamshchikov, N.V.; Prasolov, P.F.; Shestak, V.E.

    1991-01-01

    Method for determining plastic anisotropy of zurconium alloy cladding is described. It is based on consideration of material as a combination of transversal crystallites with known distribution over orientations. Such approach enables to describe cladding resistance to plastic deformation at arbitrary stressed state, using the results of texture investigations and uniaxial tests of samples, cut out of claddings along three directions. Plastic anisotropy of fuel element claddings 9.15 and 13.6 mm in diameter up to several percents of plastic deformation is shown

  5. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  6. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  7. Simplified distributed parameters BWR dynamic model for transient and stability analysis

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Nunez-Carrera, Alejandro; Vazquez-Rodriguez, Alejandro

    2006-01-01

    This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR

  8. Numerical Ballooning and Burst Prediction of Fuel Cladding During LOCA Transients in LWR

    International Nuclear Information System (INIS)

    Landau, E.; Weiss, Y.; Szanto, M.

    2014-01-01

    Modeling of nuclear fuel cladding behavior during a Loss of Coolant accident (LOCA) is a principal requirement in reactor safety analysis, most former safety criteria were obtained from experiments during the 1970's, conducted mainly with fresh fuels. Changes in modern fuel design, introduction of new cladding materials and motivation towards higher burn-ups have generated a need to re-examine safety criteria and their continued validity. This led to the growing development of both experiments and simulations meant to address this need. The Halden IFA-650 series of experiments for example, beginning in the early 2000's have clearly shown that existing criteria and experimental data are insufficient for the growing demand for higher burn-ups. Several codes for reactor core and fuel rod analysis exist nowadays, such as FRAPTRAN1.4 or RELAP5-3D . These are tailor-made codes, designed to predict general core behavior and fuel performance, and while they are also used in predicting core components behavior during accident conditions, including those of cladding ballooning and failure with good accuracy, they contain several limitations on modeling the full transient cladding thermo mechanical phenomena. Limitations such as mechanical models being one dimensional or in axisymmetric geometries only, relying mostly on analytical models therefore having further restricting assumptions in return for accuracy, etc. These limitations disable the simulation of several important aspects, such as modeling 3D azimuthal behavior for example. The objective of the current work is to develop a comprehensive numerical model for predicting zircalloy cladding thermo mechanical behavior during a LOCA. The model will eventually predicts full cladding ballooning and burst behavior followed by fuel relocation, for fuel rods that can be subjected to 3D distributed flux. The model is fully three dimensional and is created using the commercial FEM numerical simulation software ABAQUS© applying

  9. Modern Cosmology

    CERN Document Server

    Zhang Yuan Zhong

    2002-01-01

    This book is one of a series in the areas of high-energy physics, cosmology and gravitation published by the Institute of Physics. It includes courses given at a doctoral school on 'Relativistic Cosmology: Theory and Observation' held in Spring 2000 at the Centre for Scientific Culture 'Alessandro Volta', Italy, sponsored by SIGRAV-Societa Italiana di Relativita e Gravitazione (Italian Society of Relativity and Gravitation) and the University of Insubria. This book collects 15 review reports given by a number of outstanding scientists. They touch upon the main aspects of modern cosmology from observational matters to theoretical models, such as cosmological models, the early universe, dark matter and dark energy, modern observational cosmology, cosmic microwave background, gravitational lensing, and numerical simulations in cosmology. In particular, the introduction to the basics of cosmology includes the basic equations, covariant and tetrad descriptions, Friedmann models, observation and horizons, etc. The ...

  10. Modern bureaucracy

    OpenAIRE

    Toye, John

    2006-01-01

    Max Weber believed that bureaucracy could be understood by analysing its ideal-typical characteristics, and that these characteristics would become more pervasive as the modern age advanced. Weber’s horizontal account of bureaucracy can be criticised on various grounds, including its unrealistic notion of bureaucratic rationality. An alternative view is proposed, namely, that the development of state bureaucracies is driven by the trajectory of the highpower politics in which they are nested....

  11. Modern Biology

    OpenAIRE

    ALEKSIC, Branko

    2014-01-01

    The purpose of this course is to learn the philosophy, principles, and techniques of modern biology. The course is particularly designed for those who have not learned biology previously or whose major is other than biology, and who may think that they do not need to know any biology at all. The topics are covered in a rather general, overview manner, but certain level of diligence in grasping concepts and memorizing the terminology is expected.

  12. Modern maths

    CERN Multimedia

    Thom,R

    1974-01-01

    Le Prof. R. Thom expose ses vues sur l'enseignement des mathématiques modernes et des mathémathiques de toujours. Il est un grand mathématicien et était professeur à Strasbourg; maintenant il est professeur de hautes études scientifiques et était invité par le Prof. Piaget à Genève

  13. Prony's method application for BWR instabilities characterization

    Energy Technology Data Exchange (ETDEWEB)

    Castillo, Rogelio, E-mail: rogelio.castillo@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Ramírez, J. Ramón, E-mail: ramon.ramirez@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Alonso, Gustavo, E-mail: gustavo.alonso@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico); Instituto Politecnico Nacional, Unidad Profesional Adolfo Lopez Mateos, Ed. 9, Lindavista, D.F. 07300 (Mexico); Ortiz-Villafuerte, Javier, E-mail: javier.ortiz@inin.gob.mx [Instituto Nacional de Investigaciones Nucleares, Carretera México-Toluca s/n, La Marquesa, Ocoyoacac, Estado de México 52750 (Mexico)

    2015-04-01

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred.

  14. Prevention of organic iodide formation in BWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Karjunen, T [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Laitinen, T; Piippo, J; Sirkiae, P [VTT Manufacturing Technology (Finland)

    1996-12-01

    During an accident, many different forms of iodine may emerge. Organic iodides, such as methyl iodide and ethyl iodide, are relatively volatile, and thus their appearance leads to increased concentration of gaseous iodine. Since organic iodides are also relatively immune to most accident mitigation measures, such as sprays and filters, they can affect the accident source term significantly even when only a small portion of iodine is in organic form. Formation of organic iodides may not be limited by the amount of organic substances available. Excessive amounts of methane can be produced, for example, during oxidation of boron carbide, which is used in BWR`s as a neutron absorber material. Another important source is cable insulation. In a BWR, a large quantity of cables is placed below the pressure vessel. Thus a large quantity of pyrolyse gases will be produced, should the vessel fail. Organic iodides can be formed as a result of many different reactions, but at least in certain conditions the main reaction takes place between an organic radical produced by radiolysis and elemental iodine. A necessary requirement for prevention of organic iodide production is therefore that the pH in the containment water pools is kept high enough to eliminate formation of elemental iodine. In a typical BWR the suppression pool water is usually unbuffered. As a result, the pH may be dominated by chemicals introduced during an accident. If no system for adding basic chemicals is operable, the main factor affecting pool water pH may be hydrochloric acid released during cable degradation. Should this occur, the conditions could be very favorable for production of elemental iodine and, consequently, formation of organic iodides. Although high pH is necessary for iodine retention, it could have also adverse effects. High pH may, for example, accelerate corrosion of containment materials and alter the characteristics of the solid corrosion products. (author) 6 figs., 1 tab., 13 refs.

  15. Developing and modeling of the 'Laguna Verde' BWR CRDA benchmark

    International Nuclear Information System (INIS)

    Solis-Rodarte, J.; Fu, H.; Ivanov, K.N.; Matsui, Y.; Hotta, A.

    2002-01-01

    Reactivity initiated accidents (RIA) and design basis transients are one of the most important aspects related to nuclear power reactor safety. These events are re-evaluated whenever core alterations (modifications) are made as part of the nuclear safety analysis performed to a new design. These modifications usually include, but are not limited to, power upgrades, longer cycles, new fuel assembly and control rod designs, etc. The results obtained are compared with pre-established bounding analysis values to see if the new core design fulfills the requirements of safety constraints imposed on the design. The control rod drop accident (CRDA) is the design basis transient for the reactivity events of BWR technology. The CRDA is a very localized event depending on the control rod insertion position and the fuel assemblies surrounding the control rod falling from the core. A numerical benchmark was developed based on the CRDA RIA design basis accident to further asses the performance of coupled 3D neutron kinetics/thermal-hydraulics codes. The CRDA in a BWR is a mostly neutronic driven event. This benchmark is based on a real operating nuclear power plant - unit 1 of the Laguna Verde (LV1) nuclear power plant (NPP). The definition of the benchmark is presented briefly together with the benchmark specifications. Some of the cross-sections were modified in order to make the maximum control rod worth greater than one dollar. The transient is initiated at steady-state by dropping the control rod with maximum worth at full speed. The 'Laguna Verde' (LV1) BWR CRDA transient benchmark is calculated using two coupled codes: TRAC-BF1/NEM and TRAC-BF1/ENTREE. Neutron kinetics and thermal hydraulics models were developed for both codes. Comparison of the obtained results is presented along with some discussion of the sensitivity of results to some modeling assumptions

  16. Prony's method application for BWR instabilities characterization

    International Nuclear Information System (INIS)

    Castillo, Rogelio; Ramírez, J. Ramón; Alonso, Gustavo; Ortiz-Villafuerte, Javier

    2015-01-01

    Highlights: • Prony's method application for BWR instability events. • Several BWR instability benchmark are assessed using this method. • DR and frequency are obtained and a new parameter is proposed to eliminate false signals. • Adequate characterization of in-phase and out-of-phase events is obtained. • The Prony's method application is validated. - Abstract: Several methods have been developed for the analysis of reactor power signals during BWR power oscillations. Among them is the Prony's method, its application provides the DR and the frequency of oscillations. In this paper another characteristic of the method is proposed to determine the type of oscillations that can occur, in-phase or out-of-phase. Prony's method decomposes a given signal in all the frequencies that it contains, therefore the DR of the fundamental mode and the first harmonic are obtained. To determine the more dominant pole of the system a normalized amplitude W of the system is calculated, which depends on the amplitude and the damping coefficient. With this term, it can be analyzed which type of oscillations is present, if W of the fundamental mode frequency is the greater, the type of oscillations is in-phase, if W of the first harmonic frequency is the greater, the type of oscillations is out-of-phase. The method is applied to several stability benchmarks to assess its validity. Results show the applicability of the method as an alternative analysis method to determine the type of oscillations occurred

  17. FRAPCON analysis of cladding performance during dry storage operations

    Directory of Open Access Journals (Sweden)

    David J. Richmond

    2018-03-01

    Full Text Available There is an increasing need in the United States and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations or interim storage sites. Under normal conditions, the Nuclear Regulatory Commission limits cladding temperature to 400°C for high-burnup (>45 GWd/mtU fuel, with higher temperatures allowed for low-burnup fuel. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400°C. Results were representative of the majority of US light water reactor fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage. Keywords: Dry Storage, FRAPCON, Fuel Performance, Radial Hydride Reorientation, Vacuum Drying

  18. Duplex-cladding: Siemens answer to the requirements of extended burnup in PWRs

    International Nuclear Information System (INIS)

    Van Swam, L.F.; Sell, H.J.; Eberle, R.; Seibold, A.

    1994-01-01

    One important goal of nuclear fuel development is to increase the cost-effectiveness of the nuclear fuel cycle by burnup extension. A prerequisite for this goal is a cladding tube with high resistance to corrosion under the operating conditions of modern PWRs. Therefore, in the early eighties Siemens started to investigate the material behaviour of Zirconium based alloys also outside the composition range of Zry-4. The examination included out-of-pile corrosion testing in water and steam, with and without chemical addition, such as LiOH, in-pile testing of path finder fuel rods in a hot PWR up to 80 MWd/kgU and the investigation of mechanical behaviour, growth and creep under normal and the postulated conditions of a loss-of-coolant accident (LOCA). The evaluation of in-pile and out-of-pile experiments on alternative Zr-alloys revealed that improvements in corrosion resistance are frequently accompanied by undesirable changes in material properties which affect mechanical design and LOCA behaviour. To fulfill all requirements - the mechanical and corrosion related ones - and to retain the large experience base with Zry-4, a DUPLEX cladding was selected. The selected ELS DUPLEX cladding consists of a Zircaloy-4 tubing with a thin outer layer of an Extra Low tin (Sn) Zr-alloy. The ELS layer improves the stability against LiOH and allows operation with voided coolant. This advanced product has been engineered for use in highly enriched fuel assemblies in high efficiency plants operating with low neutron leakage core management and high coolant temperatures. It has become the accepted fuel rod cladding for many plants in Germany, Spain and Switzerland. (authors). 6 figs., 2 refs

  19. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    Hsiang-Shou Cheng; Diamond, D.J.

    1978-01-01

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

  20. Measurements and analysis of neutron and gamma noise in BWR's

    International Nuclear Information System (INIS)

    Dam, H. van; Kleiss, E.B.J.

    1985-01-01

    Neutron and gamma sensitive collectrons (self-powered detectors) have been designed for incore noise measurements in BWRs. A so-called twin-type has been developed for measurements of two-phase flow characteristics and detailed axial velocity distributions. Construction aspects of the twin detectors are discussed. An analysis is presented of the response of both detector types to incore parametric fluctuations. This analysis is based on detector response functions which provide an insight into the 'field of view' of the two types. The results are supported by experimental verifications; it is shown that incore gamma detectors provide useful additional information about two-phase flow in a BWR. (author)

  1. RETRAN experience with BWR transients at Yankee Atomic Electric Company

    International Nuclear Information System (INIS)

    Ansari, A.A.F.; Cronin, J.T.; Slifer, B.C.

    1981-01-01

    Yankee Atomic Electric Company is actively involved in the development of licensing methods for BWR's. The computer code chosen for analyzing system response under transient conditions is RETRAN. This paper describes the RETRAN model developed for Vermont Yankee, and the results of the RETRAN checkout and qualification that has been achieved at YAEC through comparison of RETRAN predictions to the startup test results performed at the plant as part of the 100% power startup test program. In addition, abnormal operational transients typically analyzed for licensing are also presented

  2. A study on the feasibility of minor actinides in BWR

    International Nuclear Information System (INIS)

    Abdul Waris; Budiono

    2008-01-01

    Preliminary study on the feasibility of actinides minor (MA) recycling without mixing them with plutonium in boiling water reactor (BWR) has been carried out. The results show that increasing of fissile MA content in mixed oxide fuel (MOX) and/or reducing void fraction can enlarge the effective multiplication factor at the beginning of cycle, but the reactor still can not obtain its criticality condition. Furthermore, dropping the void fraction results in higher reactivity swing and therefore plummeting the safety factor of the reactor. (author)

  3. Corrosion failure of a BWR embedded reactor containment liner

    International Nuclear Information System (INIS)

    Wegemar, B.

    2006-01-01

    Following sixteen fuel cycles, leakage through a BWR embedded reactor containment liner (carbon steel) was discovered. Leakage was located at a penetration for electrical conductors as a result of penetrating corrosion attack. During construction, porous cement structures and air pockets/cavities were formed due to erroneous injection of grout. Corrosion attacks on the CS steel liner were located at the relatively small, active surfaces in contact with the porous cement structure. The corrosion mechanism was supposed to be anodic dissolution of the steel liner in areas with insufficient passivation. The penetrations were restored according to original design requirements. (author)

  4. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

  5. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    Tavera D, L.; Camacho L, M.E.

    1991-01-01

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  6. The mechanical structure of the SVEA BWR fuel

    International Nuclear Information System (INIS)

    Nylund, O.; Johansson, A.; Junkrans, S.

    1985-01-01

    The SVEA BWR fuel assembly design is characterized by a double-wall cruciform internal structure forming an internal water gap and dividing the assembly into 4 subbundles. The effect is a favourable distribution of fuel and moderator, a minimum amount of structural material in active core, a combination of structural stability and flexibility for minimum control rod friction in reduced gaps and a reduced creep deformation of the fuel assembly. The results of a laboratory test program confirm the much lower friction force obtained with the SVEA fuel assemblies while withdrawing and inserting the control rod. (RF)

  7. In-situ testing of BWR closure head studs

    International Nuclear Information System (INIS)

    deRaad, J.A.; Wolters, J.T.

    1988-01-01

    Mechanized ultrasonic inspection of closure head studs often is on the critical path. In German BWR's, a floodcompensator is used which allows human access to the studs despite the water is up to a much higher level. For stud inspection this provides a potential solution to get out of the critical path. However, the space restrictions around the studs due to the geometry of the floodcompensator did not allow the use of the existing manipulators. This paper describes the design of a dedicated compact manipulator of a construction which copes with the restricted space available around the studs

  8. Decontamination and materials corrosion concerns in the BWR

    International Nuclear Information System (INIS)

    Gordon, B.M.; Gordon, G.M.

    1988-01-01

    The qualification of chemical decontamination processes to decontaminate complete systems or individual components in essential if effective inspection, maintenance, repair or replacement of plant components is to be achieved with minimum exposure of workers to ionizing radiation. However, it is critical that the benefits of decontamination processes are not overshadowed by deleterious materials/ corrosion side effects during the application of the process or during subsequent operation. This paper discusses such potential corrosion/materials problems in the BWR and presents relevant available corrosion data for the various commercial decontamination processes. (author)

  9. Development of water chemistry diagnosis system for BWR primary loop

    International Nuclear Information System (INIS)

    Nagase, Makoto; Asakura, Yamato; Sakagami, Masaharu; Uchida, Shunsuke; Ohsumi, Katsumi.

    1988-01-01

    The prototype of a water chemistry diagnosis system for BWR primary loop has been developed. Its purposes are improvement of water chemistry control and reduction of the work burden on plant chemistry personnel. It has three main features as follows. (1) Intensifying the observation of water chemistry conditions by variable sampling intervals based on the on-line measured data. (2) Early detection of water chemistry data trends using a second order regression curve which is calculated from the measured data, and then searching the cause of anomaly if anything (3) Diagnosis of Fe concentration in feedwater using model simulations, in order to lower the radiation level in the primary system. (author)

  10. The noise analysis and the BWR operation map

    International Nuclear Information System (INIS)

    Blazquez, J.; Ballestrin, J.

    1996-01-01

    An analytical expression for the Decay Ratio is obtained: DR = exp(-bW / P 1/2 ). The physics behind is also explained. It applies to a commercial BWR Operation Map, on the vicinity of the power instability. This functional form seems fitting to the structure of the Operation map. The power P and the coolant flow are measured straightforward; the Decay Ratio is obtained by neutron noise analysis techniques. The parameter b, depending on the void reactivity coefficient, is then calculated on line during the Reactor Operation. New DR value is now predicted for each new displacement on the Map, so unexpected instability events are more likely avoided. (authors)

  11. All fiber cladding mode stripper with uniform heat distribution and high cladding light loss manufactured by CO2 laser ablation

    Science.gov (United States)

    Jebali, M. A.; Basso, E. T.

    2018-02-01

    Cladding mode strippers are primarily used at the end of a fiber laser cavity to remove high-power excess cladding light without inducing core loss and beam quality degradation. Conventional manufacturing methods of cladding mode strippers include acid etching, abrasive blasting or laser ablation. Manufacturing of cladding mode strippers using laser ablation consist of removing parts of the cladding by fused silica ablation with a controlled penetration and shape. We present and characterize an optimized cladding mode stripper design that increases the cladding light loss with a minimal device length and manufacturing time. This design reduces the localized heat generation by improving the heat distribution along the device. We demonstrate a cladding mode stripper written on a 400um fiber with cladding light loss of 20dB, with less than 0.02dB loss in the core and minimal heating of the fiber and coating. The manufacturing process of the designed component is fully automated and takes less than 3 minutes with a very high throughput yield.

  12. DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

    International Nuclear Information System (INIS)

    Rudisill, T; John Mickalonis, J

    2006-01-01

    The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO 2 ) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH 4 F)/ammonium nitrate (NH 4 NO 3 ) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

  13. Rheological evaluation of pretreated cladding removal waste

    International Nuclear Information System (INIS)

    McCarthy, D.; Chan, M.K.C.; Lokken, R.O.

    1986-01-01

    Cladding removal waste (CRW) contains concentrations of transuranic (TRU) elements in the 80 to 350 nCi/g range. This waste will require pretreatment before it can be disposed of as glass or grout at Hanford. The CRW will be pretreated with a rare earth strike and solids removal by centrifugation to segregate the TRU fraction from the non-TRU fraction of the waste. The centrifuge centrate will be neutralized with sodium hydroxide. This neutralized cladding removal waste (NCRW) is expected to be suitable for grouting. The TRU solids removed by centrifugation will be vitrified. The goal of the Rheological Evaluation of Pretreated Cladding Removal Waste Program was to evaluate those rheological and transport properties critical to assuring successful handling of the NCRW and TRU solids streams and to demonstrate transfers in a semi-prototypic pumping environment. This goal was achieved by a combination of laboratory and pilot-scale evaluations. The results obtained during these evaluations were correlated with classical rheological models and scaled-up to predict the performance that is likely to occur in the full-scale system. The Program used simulated NCRW and TRU solid slurries. Rockwell Hanford Operations (Rockwell) provided 150 gallons of simulated CRW and 5 gallons of simulated TRU solid slurry. The simulated CRW was neutralized by Pacific Northwest Laboratory (PNL). The physical and rheological properties of the NCRW and TRU solid slurries were evaluated in the laboratory. The properties displayed by NCRW allowed it to be classified as a pseudoplastic or yield-pseudoplastic non-Newtonian fluid. The TRU solids slurry contained very few solids. This slurry exhibited the properties associated with a pseudoplastic non-Newtonian fluid

  14. PCI resistant light water reactor fuel cladding

    International Nuclear Information System (INIS)

    Foster, J.P.; Sabol, G.P.

    1988-01-01

    A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities

  15. Analysis of pellet cladding mechanical interaction using computational simulation

    Energy Technology Data Exchange (ETDEWEB)

    Berretta, José R.; Suman, Ricardo B.; Faria, Danilo P.; Rodi, Paulo A., E-mail: jose.berretta@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil); Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo (LabRisco/USP), São Paulo, SP (Brazil). Laboratório de Análise, Avaliação e Gerenciamento de Riscos

    2017-07-01

    During the operation of Pressurized Water Reactors (PWR), specifically under power transients, the fuel pellet experiences many phenomena, such as swelling and thermal expansion. These dimensional changes in the fuel pellet can enable occurrence of contact it and the cladding along the fuel rod. Thus, pellet cladding mechanical interaction (PCMI), due this contact, induces stress increase at the contact points during a period, until the accommodation of the cladding to the stress increases. This accommodation occurs by means of the cladding strain, which can produce failure, if the fuel rod deformation is permanent or the burst limit of the cladding is reached. Therefore, the mechanical behavior of the cladding during the occurrence of PCMI under power transients shall be investigated during the fuel rod design. Considering the Accident Tolerant Fuel program which aims to develop new materials to be used as cladding in PWR, one important design condition to be evaluated is the cladding behavior under PCMI. The purpose of this paper is to analyze the effects of the PCMI on a typical PWR fuel rod geometry with stainless steel cladding under normal power transients using computational simulation (ANSYS code). The PCMI was analyzed considering four geometric situations at the region of interaction between pellet and cladding. The first case, called “perfect fuel model” was used as reference for comparison. In the second case, it was considered the occurrence of a pellet crack with the loss of a chip. The goal for the next two cases was that a pellet chip was positioned into the gap of pellet-cladding, in the situations described in the first two cases. (author)

  16. Process for producing clad superconductive materials

    International Nuclear Information System (INIS)

    Cass, R.B.; Ott, K.C.; Peterson, D.E.

    1992-01-01

    This patent describes a process for fabricating superconducting composite wire. It comprises placing a superconductive precursor admixture capable of undergoing self propagating combustion in stoichiometric amounts sufficient to form a superconductive product within an oxygen-porous metal tube; sealing one end of the tube; igniting the superconductive precursor admixture whereby the superconductive precursor admixture endburns along the length of the admixture; and cross-section reducing the tube at a rate substantially equal to the rate of burning of the superconductive precursor admixture and at a point substantially planar with the burnfront of the superconductive precursor mixture, whereby a clad superconductive product is formed in situ

  17. Method and device for weld deposit cladding

    International Nuclear Information System (INIS)

    Barger, J.J.

    1977-01-01

    In order to get weld beads of good quality, uniform thickness and faultless transition regions between neighboring beads in weld deposit cladding of metallic workpoeces, it is proposed to use a device in which the electromagnets are arranged adjacent to th zone of molten welding powder and molten metal besides having got suitable supplies for applying the welding powder, the polarity of the magnets being chosen in such a way that the lines of flux between the poles are counteracting the lines of flux surrounding the electrode band because of the welding current. Several variants of arranging the electrodes are presented in detail. (UWI) [de

  18. Zircaloy cladding degradation under repository conditions

    International Nuclear Information System (INIS)

    Santanam, L.; Raghavan, S.; Chin, B.A.

    1990-12-01

    Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340 degree C (613 K) for typically stressed rods (70--100 MPa) and 300 degree C (573 K) for highly stressed rods (140--160 MPa). 10 refs., 2 figs

  19. Termination of plastic-clad fiber

    International Nuclear Information System (INIS)

    Nance, W.R.

    1982-03-01

    Optical waveguides are ideal in a nuclear weapon environment because of their resistance to electromagnetic interference. Of the fibers on today's market, plastic-clad silica (PCS) is the most radiation resistant and therfore the best choice. Because terminating PCS is complex, this paper attemps to address the major problems associated with these terminations including selecting the proper connector and optimizing the terminating procedures. The sources of losses in the connectors are summarized and typical loss values are given for four connectors which were tested

  20. CO2 laser cladding of VERSAlloyTM on carbon steel with powder feeding

    International Nuclear Information System (INIS)

    Kim, Jae-Do; Kweon, Jin-Wook

    2007-01-01

    Laser cladding processing with metal powder feeding has been experimented on carbon steel with VERSAlloy TM . A special device for the metal powder feeding was designed and manufactured. By adopting proper cladding parameters, good clad layers and sound metallurgical bonding with the base metal were obtained. Analysis indicates that the micro hardness of clad layer and the heat-affected zone increased with increasing of cladding speed. The experimental results showed that VERSAlloy TM cladded well with carbon steel

  1. Modern Cosmology

    Energy Technology Data Exchange (ETDEWEB)

    Zhang Yuanzhong

    2002-06-21

    This book is one of a series in the areas of high-energy physics, cosmology and gravitation published by the Institute of Physics. It includes courses given at a doctoral school on 'Relativistic Cosmology: Theory and Observation' held in Spring 2000 at the Centre for Scientific Culture 'Alessandro Volta', Italy, sponsored by SIGRAV-Societa Italiana di Relativita e Gravitazione (Italian Society of Relativity and Gravitation) and the University of Insubria. This book collects 15 review reports given by a number of outstanding scientists. They touch upon the main aspects of modern cosmology from observational matters to theoretical models, such as cosmological models, the early universe, dark matter and dark energy, modern observational cosmology, cosmic microwave background, gravitational lensing, and numerical simulations in cosmology. In particular, the introduction to the basics of cosmology includes the basic equations, covariant and tetrad descriptions, Friedmann models, observation and horizons, etc. The chapters on the early universe involve inflationary theories, particle physics in the early universe, and the creation of matter in the universe. The chapters on dark matter (DM) deal with experimental evidence of DM, neutrino oscillations, DM candidates in supersymmetry models and supergravity, structure formation in the universe, dark-matter search with innovative techniques, and dark energy (cosmological constant), etc. The chapters about structure in the universe consist of the basis for structure formation, quantifying large-scale structure, cosmic background fluctuation, galaxy space distribution, and the clustering of galaxies. In the field of modern observational cosmology, galaxy surveys and cluster surveys are given. The chapter on gravitational lensing describes the lens basics and models, galactic microlensing and galaxy clusters as lenses. The last chapter, 'Numerical simulations in cosmology', deals with spatial and

  2. Modern plasmonics

    CERN Document Server

    Maradudin, Alexei A; Barnes, William L

    2014-01-01

    Plasmonics is entering the curriculum of many universities, either as a stand alone subject, or as part of some course or courses. Nanotechnology institutes have been, and are being, established in universities, in which plasmonics is a significant topic of research. Modern Plasmonics book offers a comprehensive presentation of the properties of surface plasmon polaritons, in systems of different structures and various natures, e.g. active, nonlinear, graded, theoretical/computational and experimental techniques for studying them, and their use in a variety of applications. Contains materia

  3. Modern spectroscopy

    CERN Document Server

    Hollas, J Michael

    2013-01-01

    The latest edition of this highly acclaimed title introduces the reader to a wide range of spectroscopies, and includes both the background theory and applications to structure determination and chemical analysis.  It covers rotational, vibrational, electronic, photoelectron and Auger spectroscopy, as well as EXAFs and the theory of lasers and laser spectroscopy. A  revised and updated edition of a successful, clearly written book Includes the latest developments in modern laser techniques, such as cavity ring-down spectroscopy and femtosecond lasers Provides numerous worked examples, calculations and questions at the end of chapters.

  4. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L; Camacho L, M E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  5. Notes on safety modernization of nuclear power plants in Sweden

    International Nuclear Information System (INIS)

    Hammar, L.

    1997-01-01

    Investment programmes are pursued for all reactor generations in the order of 70M USD per year and unit, despite the political decision to phase out nuclear power. 15-20% of this may be safety-related. Major redesign and replacements of piping and joints in the primary system of Ringhals-1 (1:st generation BWR, external pump loops) are under-way aimed at enhancing the barrier reliability at par with with the new reactors with internal circulation pumps. Major upgrading in process control are typically on the agenda, e.g. modern digital protection and control systems as installed in Ringhals 1 and 2 in 1995. Comprehensive modernization of the control rooms are planned for all Swedish reactors, commencing in 1997 with the Forsmark reactors. The needs for modernization in regard of safety are expected to be further clarified in the design basis reviews due for completion in 1998

  6. Basic study on BWR plant behavior under the condition of severe accident

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Jyohko, Shingo; Dohgo, Hirofumi

    2015-01-01

    In this paper, we report on the results using the BWR plant simulator about the plant behavior under the condition of the severe accident that LOCA occurs but ECCS fails the water irrigation into the reactor core. The simulation experiments were carried out for the cases that LOCA has occurred in the main steam piping or in the recirculation piping, respectively. As for the results about the relationship between the LOCA area and the time from LOCA occurs until the fuel temperature rise start, the effect that RCIC operated was extremely big for LOCA area of up to 100 cm"2 for both type LOCA. In the case of main steam system LOCA, the core water level suddenly decreased for large LOCA of 2000 cm"2 area, however, if the irrigation into the reactor core was carried out 30 min after LOCA occurrence, the core had little damage. In addition, the H_2 concentration in the containment vessel did not exceed both limits of H_2 explosion nor detonation. The pressure of the containment vessel was around 3 kg/cm"2 of design value, so the soundness of the containment vessel was confirmed. On the other hand, for the recirculation system LOCA of 2000 cm"2 area, a drop of the core water level was extremely in comparison with main steam system LOCA, and the fuel assemblies were completely exposed during up to 30 min, to the irrigation from approximately 100 sec, after LOCA occurrence. Therefore, the fuel temperature during the irrigation had reached approximately 1900degC. Thus, the fuel cladding were damaged approximately less than 10%, and H_2 concentration in the containment vessel was approximately 9% which did not exceed H_2 detonation limit of 13% but exceeded H_2 explosion limit of 4%. However, the containment vessel internal pressure was settled around design pressure value of containment vessel. As the results, some core damage could not be avoided, but soundness of the containment vessel, which should take the role of 'confine', was found to be secured. (author)

  7. BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

    Energy Technology Data Exchange (ETDEWEB)

    Bevard, Bruce Balkcom [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mertyurek, Ugur [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is

  8. Long-range plasmonic waveguides with hyperbolic cladding

    DEFF Research Database (Denmark)

    Babicheva, Viktoriia E.; Shalaginov, Mikhail Y.; Ishii, Satoshi

    2015-01-01

    waveguides. We show that the proposed structures support long-range surface plasmon modes, which exist when the permittivity of the core matches the transverse effective permittivity component of the metamaterial cladding. In this regime, the surface plasmon polaritons of each cladding layer are strongly...

  9. Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

    International Nuclear Information System (INIS)

    Wilkins, S.C.

    1977-06-01

    Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

  10. Corrosion behavior of duplex and reference cladding in NPP Grohnde

    International Nuclear Information System (INIS)

    Besch, O.A.; Yagnik, S.K.; Eucken, C.M.; Bradley, E.R.

    1996-01-01

    The Nuclear Fuel Industry Research (NFIR) Group undertook a lead test assembly (LTA) program in NPP Grohnde PWR in Germany to assess the corrosion performance of duplex and reference cladding. Two identical 16 by 16 LTAs, each containing 32 peripheral test rods, completed four reactor cycles, reaching a peak rod burnup of 46 MWd/kgU. The results from poolside examinations performed at the end of each cycle, together with power histories and coolant chemistry, are reported. Five different cladding materials were characterized during fabrication. The corrosion performance of the cladding materials was tracked in long-term tests in high-pressure, high-temperature autoclaves. The relative ranking of corrosion behavior in such tests corresponded well with the in-reactor corrosion performance. The extent and distribution of hydriding in duplex and reference specimens during the autoclave testing has been characterized. The in-reactor corrosion data indicate that the low-tin Zircaloy-4 reference cladding, R2, had an improved corrosion resistance compared to high-tin Zircaloy-4 reference cladding, R1. Two types of duplex cladding, D1 (Zr-2.5% Nb) and D2 (Zr-0.4% Fe-0.5% Sn), showed an even further improvement in corrosion resistance compared to R2 cladding. The third duplex cladding, D3 (Zr-4 + 1.0% Nb), had significantly less corrosion resistance, which was inferior to R1. The in-reactor and out-reactor corrosion performances have been ranked

  11. Cladding of Advanced Al Alloys Employing Friction Stir Welding

    NARCIS (Netherlands)

    van der Stelt, A.A.; Bor, Teunis Cornelis; Geijselaers, Hubertus J.M.; Akkerman, Remko; van den Boogaard, Antonius H.

    2013-01-01

    In this paper an advanced solid state cladding process, based on Friction Stir Welding, is presented. The Friction Surface Cladding (FSC) technology enables the deposition of a solid-state coating using filler material on a substrate with good metallurgical bonding. A relatively soft AA1050 filler

  12. Efforts for optimization of BWR core internals replacement

    International Nuclear Information System (INIS)

    Iizuka, N.

    2000-01-01

    The core internal components replacement of a BWR was successfully completed at Fukushima-Daiichi Unit 3 (1F3) of the Tokyo Electric Power Company (TEPCO) in 1998. The core shroud and the majority of the internal components made by type 304 stainless steel (SS) were replaced with the ones made of low carbon type 316L SS to improve Intergranular Stress Corrosion Cracking (IGSCC) resistance. Although this core internals replacement project was completed, several factors combined to result in a longer-than-expected period for the outage. It was partly because the removal work of the internal components was delayed. Learning a lesson from whole experience in this project, some methods were adopted for the next replacement project at Fukushima-Daiichi Unit 2 (1F2) to shorten the outage and reduce the total radiation exposure. Those are new removal processes and new welding machine and so on. The core internals replacement work was ended at 1F2 in 1999, and both the period of outage and the total radiation exposure were the same degree as expected previous to starting of this project. This result shows that the methods adopted in this project are basically applicable for the core internals replacement work and the whole works about the BWR core internals replacement were optimized. The outline of the core internals replacement project and applied technologies at 1F3 and 1F2 are discussed in this paper. (author)

  13. BWR power oscillation evaluation methodologies in core design

    International Nuclear Information System (INIS)

    Hotta, Akitoshi

    1995-01-01

    At the initial stage of BWR development, the power oscillation due to the nuclear-thermal interaction originated in random boiling phenomena and nuclear void feedback was feared. But it was shown that under the high pressure condition in the normal operation of recent commercial BWRs, the core is in very stable state. However, power oscillation events have been observed in actual machines, and it is necessary to do the stability evaluation that sufficiently reflects the detailed operation conditions of actual plants. As the cause of power oscillation events, the instability of control system and nuclear-thermal coupling instability are important, and their mechanisms are explained. As the model for analyzing the stability of BWR core, the nuclear-thermal coupling model in frequency domain is the central existence. As the information for the design, the parameters of fuel assemblies, and the nuclear parameters and the thermohydraulic parameters of cores are enumerated. LAPUR-TSI is a nuclear-thermal coupling model. The analysis system in the software of Tokyo Electric Power Co. is outlined, and the analysis model was verified. (K.I.)

  14. Examination of minor actinide annihilation by BWR core

    International Nuclear Information System (INIS)

    Hida, Kazuki

    1995-01-01

    From the viewpoint of reducing burden for disposing high level waste generated from spent fuel, the examination of recycling minor actinide (MA) to reactors and reducing its accumulation has been advanced. In this study, the possibility of annihilation in the case of recycling it to a BWR was examined. The main MAs are 237 Np, 241 Am, 243 Am, 242 Cm, and 244 Cm. However, as for Cm isotopes, the half life is short, the amount of generation is small, and the rate of neutron emission is high, therefore, those are disposed as waste, and 237 Np, 241 Am and 243 Am were taken as the objects of recycling. In order to grasp the basic characteristics in the case of recycling MAs to a BWR, MAs were added to UO 2 fuel, MOX fuel and HCR fuel and burned, and the nuclear conversion characteristics were examined. As the result, it was found that they were converted to short half life nuclides, and as the neutron spectra were softer, the rate of annihilation was higher. In the case of recycling MAs by concentrating to a specific reactor, reactivity loss, the degree of uranium enrichment required for compensating reactivity, and the rate of MA annihilation were calculated. Based on these data, the MA recycling system was set up, and the rate of MA annihilation was evaluated. This is reported. (K.I.)

  15. BWR containments license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Smith, S.; Gregor, F.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

  16. Appraisal of BWR plutonium burners for energy centers

    International Nuclear Information System (INIS)

    Williamson, H.E.

    1976-01-01

    The design of BWR cores with plutonium loadings beyond the self-generation recycle (SGR) level is investigated with regard to their possible role as plutonium burners in a nuclear energy center. Alternative plutonium burner approaches are also examined including the substitution of thorium for uranium as fertile material in the BWR and the use of a high-temperature gas reactor (HTGR) as a plutonium burner. Effects on core design, fuel cycle facility requirements, economics, and actinide residues are considered. Differences in net fissile material consumption among the various plutonium-burning systems examined were small in comparison to uncertainties in HTGR, thorium cycle, and high plutonium-loaded LWR technology. Variation in the actinide content of high-level wastes is not likely to be a significant factor in determining the feasibility of alternate systems of plutonium utilization. It was found that after 10,000 years the toxicity of actinide high-level wastes from the plutonium-burning fuel cycles was less than would have existed if the processed natural ores had not been used for nuclear fuel. The implications of plutonium burning and possible future fuel cycle options on uranium resource conservation are examined in the framework of current ERDA estimates of minable uranium resources

  17. BWROPT: A multi-cycle BWR fuel cycle optimization code

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, Keith E.; Maldonado, G. Ivan, E-mail: Ivan.Maldonado@utk.edu

    2015-09-15

    Highlights: • A multi-cycle BWR fuel cycle optimization algorithm is presented. • New fuel inventory and core loading pattern determination. • The parallel simulated annealing algorithm was used for the optimization. • Variable sampling probabilities were compared to constant sampling probabilities. - Abstract: A new computer code for performing BWR in-core and out-of-core fuel cycle optimization for multiple cycles simultaneously has been developed. Parallel simulated annealing (PSA) is used to optimize the new fuel inventory and placement of new and reload fuel for each cycle considered. Several algorithm improvements were implemented and evaluated. The most significant of these are variable sampling probabilities and sampling new fuel types from an ordered array. A heuristic control rod pattern (CRP) search algorithm was also implemented, which is useful for single CRP determinations, however, this feature requires significant computational resources and is currently not practical for use in a full multi-cycle optimization. The PSA algorithm was demonstrated to be capable of significant objective function reduction and finding candidate loading patterns without constraint violations. The use of variable sampling probabilities was shown to reduce runtime while producing better results compared to using constant sampling probabilities. Sampling new fuel types from an ordered array was shown to have a mixed effect compared to random new fuel type sampling, whereby using both random and ordered sampling produced better results but required longer runtimes.

  18. Managing the aging of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.; Farmer, W.S.

    1992-01-01

    This Phase I Nuclear Plant Aging Research (NPAR) study examines the aging phenomena associated with BWR control and rod drive mechanisms (CRDMs) and assesses the merits of various methods of ''imaging'' this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of the Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the control rod drive (CRD) system, and (4) personal information exchange with nuclear industry CRDM maintenance experts. The report documenting the findings of this research, NUREG-5699, will be published this year. Nearly 23% of the NPRDS CRD system component failure reports were attributed to the CRDM. The CRDM components most often requiring replacement due to aging are the Graphitar seals. The predominant causes of aging for these seals are mechanical wear and thermal embrittlement. More than 59% of the NPRDS CRD system failure reports were attributed to components that comprise the hydraulic control unit (HCU). The predominant HCU components experiencing the effects of service wear and aging are value seals, discs, seats, stems, packing, and diaphragms

  19. Aging assessment of BWR control rod drive systems

    International Nuclear Information System (INIS)

    Greene, R.H.

    1991-01-01

    This study examines the aging phenomena associated with boiling water reactor (BWR) control rod drive mechanisms (CRDMs) and assess the merits of various methods of managing this aging. Information for this study was acquired from (1) the results of a special CRDM aging questionnaire distributed to each US BWR utility, (2) a first-of-its-kind workshop held to discuss CRDM aging and maintenance concerns, (3) an analysis of Nuclear Plant Reliability Data System (NPRDS) failure cases attributed to the CRD system, and (4) personal information exchange with industry experts. As part of this study, nearly 3500 NPRDS failure reports have been analyzed to examine the prevailing failure trends for CRD system components. An investigation was conducted to summarize the occurrence frequency of these component failures, discovery methods, reported failure causes, their respective symptoms, and actions taken by utilities to restore component and system service. The results of this research have identified the predominant CRDM failure modes and causes. In addition, recommendations are presented that identify specific actions utilities can implement to mitigate CRDM aging. An evaluation has also been made of certain maintenance practices and tooling which have enabled some utilities to reduce ALARA exposures received from routine CRDM replacement and rebuilding activities. 5 refs., 8 figs., 2 tabs

  20. Sophistication of operator training using BWR plant simulator

    International Nuclear Information System (INIS)

    Ohshiro, Nobuo; Endou, Hideaki; Fujita, Eimitsu; Miyakita, Kouji

    1986-01-01

    In Japanese nuclear power stations, owing to the improvement of fuel management, thorough maintenance and inspection, and the improvement of facilities, high capacity ratio has been attained. The thorough training of operators in nuclear power stations also contributes to it sufficiently. The BWR operator training center was established in 1971, and started the training of operators in April, 1974. As of the end of March, 1986, more than 1800 trainees completed training. At present, in the BWR operator training center, No.1 simulator of 800 MW class and No.2 simulator of 1100 MW class are operated for training. In this report, the method, by newly adopting it, good result was obtained, is described, that is, the method of introducing the feeling of being present on the spot into the place of training, and the new testing method introduced in retraining course. In the simulator training which is apt to place emphasis on a central control room, the method of stimulating trainees by playing the part of correspondence on the spot and heightening the training effect of multiple monitoring was tried, and the result was confirmed. The test of confirmation on the control board was added. (Kako, I.)