WorldWideScience

Sample records for water reactors flux

  1. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  2. Self-adjointness of the fast flux in a pressurized water reactor

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1985-01-01

    Most computer codes for the analysis of systems transients rely on a simplified representation of the active core, typically employing either a one-dimensional or a point kinetics model. The collapsing of neutronics data from multidimensional steady-state calculations normally employs flux/flux-adjoint weighting. The multidimensional calculations, however, usually are performed only for the forward problem, not the adjoint. The collapsing methodologies employed in generating the neutronics input for transient codes typically construct adjoint fluxes from the assumption that the fast flux is self-adjoint. Until now, no further verification of this assumption has been undertaken for thermal reactors. As part of the verification effort for EPRI's reactor analysis support package, the validity of this assumption now has been investigated for a modern pressurized water reactor (PWR). The PDQ-7 code was employed to perform two-group fine-mesh forward and adjoint calculations for a two-dimensional representation of Zion Unit 2 at beginning of life, based on the standard PWR ARMP model. It has been verified that the fast flux is very nearly self-adjoint in a PWR. However, a significant error can arise during the subsequent construction of the thermal adjoint flux unless allowance is made for the difference between the forward and adjoint thermal buckling terms. When such a difference is included, the thermal adjoint flux can be estimated very accurately

  3. Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

    International Nuclear Information System (INIS)

    Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

    2005-01-01

    The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

  4. Fast neutron flux in heavy water reactors; Flux de neutrons rapides dans les piles a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Brisbois, J; Katz, S [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

    1966-07-01

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author) [French] La possibilite de calculer le flux de neutrons rapides dans un reseau d'uranium naturel a eau lourde par superposition des apports des divers barreaux, a ete verifiee en utilisant un code Monte-Carlo monodimensionel. Les resultats obtenus concordent avec des mesures experimentales effectuees dans le coeur et reacteur de la pile Aquilon. (auteurs)

  5. Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

    2017-05-15

    Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

  6. Effects of some structural materials on the reactivity and flux distributions in a pressurised water reactor

    International Nuclear Information System (INIS)

    Mondal, A.M.W.; Mannan, M.A.

    1983-01-01

    The effect of the structural materials of the guide tubes, spacer grids and the shroud on the reactivity and the flux distribution of a Pressurised Water Reactor (PWR) has been studied. Group constants of different cells of guide tubes, spacer grids, shroud and the fuel have been calculated using the cell codes LEOPARD, PANTHER and METHUSELAH. Core calculations have been performed using the diffusion code EQUIPOISE. It has been found for a PWR of 1300 MWe of Kraftwork Union design for Iron that the total change in reactivity due to the presence of guide tubes, spacer grids and the shroud is about -2.48x10 -2 . (author)

  7. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR)

    Science.gov (United States)

    Brovchenko, Mariya; Dechenaux, Benjamin; Burn, Kenneth W.; Console Camprini, Patrizio; Duhamel, Isabelle; Peron, Arthur

    2017-09-01

    The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR). The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  8. Neutron-gamma flux and dose calculations in a Pressurized Water Reactor (PWR

    Directory of Open Access Journals (Sweden)

    Brovchenko Mariya

    2017-01-01

    Full Text Available The present work deals with Monte Carlo simulations, aiming to determine the neutron and gamma responses outside the vessel and in the basemat of a Pressurized Water Reactor (PWR. The model is based on the Tihange-I Belgian nuclear reactor. With a large set of information and measurements available, this reactor has the advantage to be easily modelled and allows validation based on the experimental measurements. Power distribution calculations were therefore performed with the MCNP code at IRSN and compared to the available in-core measurements. Results showed a good agreement between calculated and measured values over the whole core. In this paper, the methods and hypotheses used for the particle transport simulation from the fission distribution in the core to the detectors outside the vessel of the reactor are also summarized. The results of the simulations are presented including the neutron and gamma doses and flux energy spectra. MCNP6 computational results comparing JEFF3.1 and ENDF-B/VII.1 nuclear data evaluations and sensitivity of the results to some model parameters are presented.

  9. A critical heat flux correlation for advanced pressurized light water reactor application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Hame, W.

    1982-05-01

    Many CHF-correlations have been developed for water cooled rod clusters representing typical PWR or BWR fuel element geometries with relative wide rod lattices. However the fuel elements of an Advanced Pressurized Water Reactor (APWR) have a tight fuel rod lattice, in view of increasing the fuel utilization. It was therefore decided to produce a new CHF-correlation valid for rod bundles with tight lattices. The already available WSC-2 correlation was chosen as a basis. The geometry dependent parameters of this correlation were determined again with the method of the root mean square fitting from the experimental data of the CHF-tests performed in the frame of the Light Water Breeder Reactor programme at the Bettis Laboratory. These tests include triangular array rod bundles with very tight lattices. Furthermore the effect of spiral spacer ribs was investigated on the basis of experimental data from the Columbia University. Application of the new CHF-correlation to conditions typical for an APWR shows that the predicted critical heat fluxes are much smaller than those calculated with the usual PWR-CHF-correlations, but they are higher than those predicted by the B+W-VPI+SU correlation. (orig.) [de

  10. Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

    International Nuclear Information System (INIS)

    Bultman, J.

    1992-04-01

    Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

  11. Critical heat flux and flow instability in an advanced light water reactor

    International Nuclear Information System (INIS)

    Dae-Hyun Hwang; Kyong-Won Seo; Chung-Chan Lee; Sung-Kyun Zee

    2005-01-01

    Full text of publication follows: An advanced light water reactor concept has been continuously studied in KAERI with an output in the range of about 60 to 300 MW th . The reactor is purposed to be utilized as an energy source for seawater desalination as well as small scale power generation. In order to achieve the intrinsic safety and enhanced operational flexibility, some specific design considerations such as low power density and soluble boron free operation have been incorporated in the multiple-parallel-channel type reactor core. The low power density can be achieved by adopting fuel assemblies with tightly spaced non-square lattice rod array. The allowable core operating region should be primarily limited by the two design parameters; the critical heat flux(CHF) and the flow instabilities in the multiple parallel fuel assembly channels. The characteristics of CHF and flow instability have been investigated through experimental and analytical works. The CHF prediction model was established on the basis of experimental data obtained from 19-rod test bundles. The CHF experiments have been conducted for various test bundles with different heated lengths, uniform and non-uniform radial and axial power distributions, water and Freon as the working fluids, and different number of unheated rods. The parametric ranges of CHF experiments covers the pressure from 6 to 18 MPa, the mass flux from 150 to 2000 kg/m 2 /s, and the inlet subcooling from 10 to 120 deg. C. The flow instabilities due to density wave oscillations were investigated by conducting experiments with two parallel channels under the pressure ranges from 6 to 16 MPa. The parametric behavior of flow instability was examined for the test sections with different lengths of adiabatic risers, different axial power shapes, different inlet restrictions, and different channel cross sections. The stability boundary was experimentally determined by increasing channel inlet temperature or reducing the flow rate

  12. Critical heat flux experiments for high conversion light water reactor, (3)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

    1990-03-01

    As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

  13. Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

  14. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

    2017-06-15

    Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

  15. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  16. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  17. Drift flux formulation of a boiling water reactor channel with subcooled boiling

    International Nuclear Information System (INIS)

    Elias, E.E.; Shak, D.P.; May, R.S.

    1987-01-01

    The channel formulation used in the BWR module of the Modular Modeling System MMS-02 is presented. The purpose of channel model is to accurately predict the transient response of the enthalpy void and flow rate. Accurate prediction of the two-phase enthalpy, and void fraction distributions along the channel is important since they are key input parameters to the neutronic model, and have direct effect on the core and overall reactor response. In order to model the channel response correctly, the physical phenomena had to be realistically represented. The model accounts for subcooled boiling and slip through the use of an empirical subcooled void-quality model. Simplifying assumptions are made so that only one differential equation, the energy equation, is integrated along the channel. A consistent use of semi-empirical correlations enables a complete representation of the channel flow and void fraction with the bulk enthalpy as the only state variable. The differential equation and the constitutive relations of this two-phase flow model are presented. Several numerical examples are given, and finally, come conclusions are presented

  18. Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

    International Nuclear Information System (INIS)

    Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

    2016-01-01

    At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

  19. Optimization of neutron flux distribution in Isotope Production Reactor

    International Nuclear Information System (INIS)

    Valladares, G.L.

    1988-01-01

    In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely the Vmod / Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO 2 ) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author) [pt

  20. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

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

  1. High Flux Isotope Reactor technical specifications

    International Nuclear Information System (INIS)

    1985-11-01

    This report gives technical specifications for the High Flux Isotope Reactor (HFIR) on the following: safety limits and limiting safety system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls

  2. Two-dimensional DORT discrete ordinates X-Y geometry neutron flux calculations for the Halden Heavy Boiling Water Reactor core configurations

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.

    1990-07-01

    Results are reported for two-dimensional discrete ordinates, X-Y geometry calculations performed for seven Halden Heavy Boiling Water Reactor core configurations. The calculations were performed in support of an effort to reassess the neutron fluence received by the reactor vessel. Nickel foil measurement data indicated considerable underprediction of fluences by the previously used multigroup removal- diffusion method. Therefore, calculations by a more accurate method were deemed appropriate. For each core configuration, data are presented for (1) integral fluxes in the core and near the vessel wall, (2) neutron spectra at selected locations, (3) isoflux contours superimposed on the geometry models, (4) plots of the geometry models, and (5) input for the calculations. The initial calculations were performed with several mesh sizes. Comparisons of the results from these calculations indicated that the uncertainty in the calculated fluxes should be less than 10%. However, three-dimensional effects (such as axial asymmetry in the fuel loading) could contribute to much greater uncertainty in the calculated neutron fluxes. 7 refs., 22 figs., 11 tabs.

  3. Heat transfer for ultrahigh flux reactor

    International Nuclear Information System (INIS)

    Wadkins, R.P.; Lake, J.A.; Oh, C.H.

    1987-01-01

    The use of a uniquely designed nuclear reactor to supply neutrons for materials research is the focus of recent reactor design efforts. The biological, materials, and fundamental physics aspects of research require neutron fluxes much higher than present research and testing facilities can produce. The most advanced research using neutrons as probing detectors is being done in the High Flux Reactor at the Institut Laue Langeuin, France. The design of a reactor that can produce neutron fluxes of 1.0 x 10 16 n/cm 2 .s requires a relatively high power (300 MW range) and a small core volume (approximately 30 liters). This combination of power and volume leads to a high power density which places increased demands on thermal hydraulic margins

  4. Neutron flux distribution forecasting device of reactor

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi

    1991-01-01

    A neutron flux distribution is forecast by using current data obtained from a reactor. That is, the device of the present invention comprises (1) a neutron flux monitor disposed in various positions in the reactor, (2) a forecasting means for calculating and forecasting a one-dimensional neutron flux distribution relative to imaginable events by using data obtained from the neutron flux monitor and physical models, and (3) a display means for displaying the results forecast in the forecasting means to a reactor operation console. Since the forecast values for the one-dimensional neutron flux distribution relative to the imaginable events are calculated in the device of the present invention by using data obtained from the neutron flux monitor and the physical models, the data as a base of the calculation are new and the period for calculating the forecast values can be shortened. Accordingly, although there is a worry of providing some errors in the forecast values, they can be utilized sufficiently as reference data. As a result, the reactor can be operated more appropriately. (I.N.)

  5. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  6. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

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

  7. Method for optimum determination of adjustable parameters in the boiling water reactor core simulator using operating data on flux distribution

    International Nuclear Information System (INIS)

    Kiguchi, T.; Kawai, T.

    1975-01-01

    A method has been developed to optimally and automatically determine the adjustable parameters of the boiling water reactor three-dimensional core simulator FLARE. The steepest gradient method is adopted for the optimization. The parameters are adjusted to best fit the operating data on power distribution measured by traversing in-core probes (TIP). The average error in the calculated TIP readings normalized by the core average is 0.053 at the rated power. The k-infinity correction term has also been derived theoretically to reduce the relatively large error in the calculated TIP readings near the tips of control rods, which is induced by the coarseness of mesh points. By introducing this correction, the average error decreases to 0.047. The void-quality relation is recognized as a function of coolant flow rate. The relation is estimated to fit the measured distributions of TIP reading at the partial power states

  8. Neutron flux measurements in PUSPATI Triga Reactor

    International Nuclear Information System (INIS)

    Gui Ah Auu; Mohamad Amin Sharifuldin Salleh; Mohamad Ali Sufi.

    1983-01-01

    Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm 2 sec and 1.10E+12n/cm 2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

  9. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-01-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  10. Measurements of neutron flux in the RA reactor

    International Nuclear Information System (INIS)

    Raisic, N.

    1961-12-01

    This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire [sr

  11. Neutron flux enhancement in the NRAD reactor

    International Nuclear Information System (INIS)

    Weeks, A.A.; Heidel, C.C.; Imel, G.R.

    1988-01-01

    In 1987 a series of experiments were conducted at the NRAD reactor facility at Argonne National Laboratory - West (ANL-W) to investigate the possibility of increasing the thermal neutron content at the end of the reactor's east beam tube through the use of hydrogenous flux traps. It was desired to increase the thermal flux for a series of experiments to be performed in the east radiography cell, in which the enhanced flux was required in a relatively small volume. Hence, it was feasible to attempt to focus the cross section of the beam to a smaller area. Two flux traps were constructed from unborated polypropylene and tested to determine their effectiveness. Both traps were open to the entire cross-sectional area of the neutron beam (as it emerges from the wall and enters the beam room). The sides then converged such that at the end of the trap the beam would be 'focused' to a greater intensity. The differences in the two flux traps were primarily in length, and hence angle to the beam as the inlet and outlet cross-sectional areas were held constant. The experiments have contributed to the design of a flux trap in which a thermal flux of nearly 10 9 was obtained, with an enhancement of 6.61

  12. Dissecting Reactor Antineutrino Flux Calculations

    Science.gov (United States)

    Sonzogni, A. A.; McCutchan, E. A.; Hayes, A. C.

    2017-09-01

    Current predictions for the antineutrino yield and spectra from a nuclear reactor rely on the experimental electron spectra from 235U, 239Pu, 241Pu and a numerical method to convert these aggregate electron spectra into their corresponding antineutrino ones. In the present work we investigate quantitatively some of the basic assumptions and approximations used in the conversion method, studying first the compatibility between two recent approaches for calculating electron and antineutrino spectra. We then explore different possibilities for the disagreement between the measured Daya Bay and the Huber-Mueller antineutrino spectra, including the 238U contribution as well as the effective charge and the allowed shape assumption used in the conversion method. We observe that including a shape correction of about +6 % MeV-1 in conversion calculations can better describe the Daya Bay spectrum. Because of a lack of experimental data, this correction cannot be ruled out, concluding that in order to confirm the existence of the reactor neutrino anomaly, or even quantify it, precisely measured electron spectra for about 50 relevant fission products are needed. With the advent of new rare ion facilities, the measurement of shape factors for these nuclides, for many of which precise beta intensity data from TAGS experiments already exist, would be highly desirable.

  13. High Flux Materials Testing Reactor (HFR), Petten

    International Nuclear Information System (INIS)

    1975-09-01

    After conversion to burnable poison fuel elements, the High Flux Materials Testing Reactor (HFR) Petten (Netherlands), operated through 1974 for 280 days at 45 MW. Equipment for irradiation experiments has been replaced and extended. The average annual occupation by experiments was 55% as compared to 38% in 1973. Work continued on thirty irradiation projects and ten development activities

  14. maximum neutron flux at thermal nuclear reactors

    International Nuclear Information System (INIS)

    Strugar, P.

    1968-10-01

    Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr

  15. Maximum neutron flux in thermal reactors

    International Nuclear Information System (INIS)

    Strugar, P.V.

    1968-12-01

    Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

  16. Neutron flux enhancement in the NRAD reactor

    International Nuclear Information System (INIS)

    Weeks, A.A.; Heidel, C.C.; Imel, G.R.

    1988-01-01

    In 1987 a series of experiments were conducted at the NRAD reactor facility at Argonne National Laboratory - West (ANL-W) to investigate the possibility of increasing the thermal neutron content at the end of the reactor's east beam tube through the use of hydrogenous flux traps. It was desired to increase the thermal flux for a series of experiments to be performed in the east radiography cell, in which the enhanced flux was required in a relatively small volume. Hence, it was feasible to attempt to focus the cross section of the beam to a smaller area. Two flux traps were constructed from unborated polypropylene and tested to determine their effectiveness. Both traps were open to the entire cross-sectional area of the neutron beam (as it emerges from the wall and enters the beam room). The sides then converged such that at the end of the trap the beam would be 'focused' to a greater intensity. The differences in the two flux traps were primarily in length, and hence angle to the beam as the inlet and outlet cross-sectional areas were held constant. It should be noted that merely placing a slab of polypropylene in the beam will not yield significant multiplication as neutrons are primarily scattered away

  17. Super critical water reactors

    International Nuclear Information System (INIS)

    Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

    2005-01-01

    Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

  18. High Flux Isotope Reactor power upgrade status

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Hale, R.E.; Cheverton, R.D.

    1997-01-01

    A return to 100-MW operation is being planned for the High Flux Isotope Reactor (HFIR). Recent improvements in fuel element manufacturing procedures and inspection equipment will be exploited to reduce hot spot and hot streak factors sufficiently to permit the power upgrade without an increase in primary coolant pressure. Fresh fuel elements already fabricated for future use are being evaluated individually for power upgrade potential based on their measured coolant channel dimensions

  19. Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2011-05-27

    The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

  20. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  1. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  2. Efficient computations of three-dimensional fluid-structure interactions during blowdown of a pressurized water reactor - FLUX

    International Nuclear Information System (INIS)

    Schumann, U.

    1979-01-01

    A numerical method (computer program FLUX) for investigation of the loads on pressure vessel internal structures during a loss-of-coolant accident of a PWR is described. In particular, the deformation of the core barrel are determined. Under operating conditions the core barrel controls the flow path in the vessel and consists mainly out of a relatively thin cylindrical shell. (orig./HP) [de

  3. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  4. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  5. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

  6. Neutron flux measuring system for nuclear reactor

    International Nuclear Information System (INIS)

    Aoki, Kazuo.

    1977-01-01

    Purpose: To avoid the generation of an undesired scram signal due to abrupt changes in the neutron level given to the detectors disposed near the boundary between the moderator and the atmosphere. Constitution: In a nuclear reactor adapted to conduct power control by the change of the level in the moderator such as heavy water, the outputs from the neutron detectors disposed vertically are averaged and the nuclear reactor is scramed corresponding to the averaged value. In this system, moderator level detectors are additionally provided to the nuclear reactor and their outputs, moderator level signal, are sent to a power averaging device where the output signals of the neutron detectors are judged if they are delivered from neutrons in the moderator or not depending on the magnitude of the level signal and the outputs of the detectors out of the moderator are substantially excluded. The reactor interlock signal from the device is utilized as a scram signal. (Seki, T.)

  7. HFBR handbook, 1992: High flux beam reactor

    International Nuclear Information System (INIS)

    Axe, J.D.; Greenberg, R.

    1992-10-01

    Welcome to the High Flux Beam Reactor (HFBR), one of the world premier neutron research facilities. This manual is intended primarily to acquaint outside users (and new Brookhaven staff members) with (almost) everything they need to know to work at the HFBR and to help make the stay at Brookhaven pleasant as well as profitable. Safety Training Programs to comply with US Department of Energy (DOE) mandates are in progress at BNL. There are several safety training requirements which must be met before users can obtain unescorted access to the HFBR. The Reactor Division has prepared specific safety training manuals which are to be sent to experimenters well in advance of their expected arrival at BNL to conduct experiments. Please familiarize yourself with this material and carefully pay strict attention to all the safety and security procedures that are in force at the HFBR. Not only your safety, but the continued operation of the facility, depends upon compliance

  8. Surveillance programme and upgrading of the High Flux Reactor Petten

    International Nuclear Information System (INIS)

    Bieth, Michel

    1995-01-01

    The High Flux Reactor (HFR) at Petten (The Netherlands), a 45 MW light water cooled and moderated research reactor in operation during more than 30 years, has been kept up to date by replacing ageing components. In 1984, the HFR was shut down for replacement of the aluminium. reactor vessel which had been irradiated during more than 20 years. The demonstration that the new vessel contains no critical defect requires knowledge of the material properties of the aluminium alloy Al 5154 with and without neutron irradiation and of the likely defect presence through the periodic in-service inspections. An irradiation damage surveillance programme has been started in 1985 for the new vessel material to provide information on fracture mechanics properties. After the vessel replacement, the existing process of continuous upgrading and replacement of ageing components was accelerated. A stepwise upgrade of the control room is presently under realization. (author)

  9. Surveillance programme and upgrading of the High Flux Reactor Petten

    Energy Technology Data Exchange (ETDEWEB)

    Bieth, Michel [Commission of the European Communities, Joint Research Centre, Institute for Advanced Materials, High Flux Reactor Unit, Petten (Netherlands)

    1995-07-01

    The High Flux Reactor (HFR) at Petten (The Netherlands), a 45 MW light water cooled and moderated research reactor in operation during more than 30 years, has been kept up to date by replacing ageing components. In 1984, the HFR was shut down for replacement of the aluminium. reactor vessel which had been irradiated during more than 20 years. The demonstration that the new vessel contains no critical defect requires knowledge of the material properties of the aluminium alloy Al 5154 with and without neutron irradiation and of the likely defect presence through the periodic in-service inspections. An irradiation damage surveillance programme has been started in 1985 for the new vessel material to provide information on fracture mechanics properties. After the vessel replacement, the existing process of continuous upgrading and replacement of ageing components was accelerated. A stepwise upgrade of the control room is presently under realization. (author)

  10. Upgrading and modernization of the high flux reactor Petten

    International Nuclear Information System (INIS)

    Ahlf, J.

    1992-01-01

    The High Flux Reactor (HFR) at Petten, The Netherlands, owned by the European Communities and operated by the Netherlands Energy Research Foundation, is a water-cooled and moderated, multipurpose research reactor of the closed-tank in-pool type, operated at 45 MW. Performance upgrading comprised two power increases from 20 MW via 30 MW to 45 MW, providing more and higher rated irradiation positions in the tank. With the replacement of the original reactor vessel the experimental capabilities of the reactor were improved. Better pool side facilities and the introduction of a large cross-section, double, beam tube were implemented. Additional major installation upgrading activities consisted of the replacement of the primary and the pool heat exchangers, replacement of the beryllium reflector elements, extension of the overpower protection systems and upgrading of the nuclear instrumentation as well as the guaranteed power supply. Control room upgrading is in progress. A full new safety analysis, as well as the introduction of a comprehensive Quality Assurance system, are summarized under software upgrading. Continuous modernization and upgrading also takes place of equipment for fuel and structural materials irradiations for fission reactors and future fusion machines. In parallel, all supporting services, as well as the management structure for large irradiation programmes, have been developed. Presently the reactor is operating at about 275 full power days per year with an average utilization of the irradiation positions of 70 to 80%. (orig.)

  11. Reactor water chemistry control

    International Nuclear Information System (INIS)

    Kundu, A.K.

    2010-01-01

    Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

  12. New flux detectors for CANDU 6 reactors

    International Nuclear Information System (INIS)

    Cuttler, J.M.; Medak, N.

    1992-06-01

    CANDU reactors utilize large numbers of in-core self-powered detectors for control and protection. In the original design, the detectors (coaxial cables) were wound on carrier tubes and immersed in the heavy water moderator. Failures occurred due to corrosion and other factors, and replacement was very costly because the assemblies were not designed with maintenance in mind. A new design was conceived based on straight detectors, of larger diameter, in a sealed package of individual 'well' tubes. This protected the detectors from hostile environments and enabled individual failed sensors to be replaced by inserting spares in vacant neighbouring tubes. The new design was made retrofittable to older CANDU reactors. Provision was made for on-line scanning of the core with a miniature fission chamber. The modified detectors were tested in a lengthy development program and found to exhibit superior performance to that of the original detectors. Most of the CANDU reactors have now adopted the new design. In the case of the Gentilly-2 and Point Lepreau reactors, advantage was taken of the opportunity to redesign the detector layout (using better codes and the increased flexibility in positioning detectors) to achieve better coverage of abnormal events, leading to higher trip setpoints and wider operating margins

  13. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  14. Devices for irradiation of materials in the fast neutron flux at the RA heavy water reactor in Vinca; Uredjaji za ozracivanje materijala u fluksu brzih neutrona na teskovodnom reaktoru RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)

    1964-07-01

    Full text: This paper covers the concept, technical description and problems of constructing special experimental facilities at the RA reactor in Vinca. During 1962, construction materials (graphite, Mg, Al-oxides, steel and Zircaloy) were irradiated in these facilities at temperatures below 100 deg C by the integral fast neutron flux of 2 10{sup 20} n/cm{sup 2}. Temperature of the samples cooled by heavy water circulating through hollow uranium elements was measured by thermocouples. Construction of the facility, sample preparation, related measurements of the flux and irradiation of the samples were done in cooperation with the Nuclear center in Saclay, France. A short review of the experiences in using the new experimental space at the RA reactor is given, together with some problems related to transport of radioactive capsules containing samples from Vinca to Saclay.

  15. The High Flux Reactor Petten, present status and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Ahlf, J [Institute for Advanced Materials, Joint Research Centre, Petten (Netherlands)

    1990-05-01

    The High Flux Reactor (HFR) in Petten, The Netherlands, is a light water cooled and moderated multipurpose research reactor of the closed-tank in pool type. It is operated with highly enriched Uranium fuel at a power of 45 MW. The reactor is owned by the European Communities and operated under contract by the Dutch ECN. The HFR programme is funded by The Netherlands and Germany, a smaller share comes from the specific programmes of the Joint Research Centre (JRC) and from third party contract work. Since its first criticality in 1961 the reactor has been continuously upgraded by implementing developments in fuel element technology and increasing the power from 20 MW to the present 45 MV. In 1984 the reactor vessel was replaced by a new one with an improved accessibility for experiments. In the following years also other ageing equipment has been replaced (primary heat exchangers, pool heat exchanger, beryllium reflector elements, nuclear and process instrumentation, uninterruptable power supply). Control room upgrading is under preparation. A new safety analysis is near to completion and will form the basis for a renewed license. The reactor is used for nuclear energy related research (structural materials and fuel irradiations for LWR's, HTR's and FBR's, fusion materials irradiations). The beam tubes are used for nuclear physics as well as solid state and materials sciences. Radioisotope production at large scale, processing of gemstones and silicon with neutrons, neutron radiography and activation analysis are actively pursued. A clinical facility for boron neutron capture therapy is being designed at one of the large cross section beam tubes. It is foreseen to operate the reactor at least for a further decade. The exploitation pattern may undergo some changes depending on the requirements of the supporting countries and the JRC programmes. (author)

  16. The High Flux Reactor Petten, present status and prospects

    International Nuclear Information System (INIS)

    Ahlf, J.

    1990-01-01

    The High Flux Reactor (HFR) in Petten, The Netherlands, is a light water cooled and moderated multipurpose research reactor of the closed-tank in pool type. It is operated with highly enriched Uranium fuel at a power of 45 MW. The reactor is owned by the European Communities and operated under contract by the Dutch ECN. The HFR programme is funded by The Netherlands and Germany, a smaller share comes from the specific programmes of the Joint Research Centre (JRC) and from third party contract work. Since its first criticality in 1961 the reactor has been continuously upgraded by implementing developments in fuel element technology and increasing the power from 20 MW to the present 45 MV. In 1984 the reactor vessel was replaced by a new one with an improved accessibility for experiments. In the following years also other ageing equipment has been replaced (primary heat exchangers, pool heat exchanger, beryllium reflector elements, nuclear and process instrumentation, uninterruptable power supply). Control room upgrading is under preparation. A new safety analysis is near to completion and will form the basis for a renewed license. The reactor is used for nuclear energy related research (structural materials and fuel irradiations for LWR's, HTR's and FBR's, fusion materials irradiations). The beam tubes are used for nuclear physics as well as solid state and materials sciences. Radioisotope production at large scale, processing of gemstones and silicon with neutrons, neutron radiography and activation analysis are actively pursued. A clinical facility for boron neutron capture therapy is being designed at one of the large cross section beam tubes. It is foreseen to operate the reactor at least for a further decade. The exploitation pattern may undergo some changes depending on the requirements of the supporting countries and the JRC programmes. (author)

  17. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  18. Reactor water level control device

    International Nuclear Information System (INIS)

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

  19. Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell.

  20. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  1. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  2. Reactor water injection facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  3. Use of sup(233)U for high flux reactors

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Liem, P.H.

    1991-01-01

    The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

  4. Radiation dosimetry at the BNL High Flux Beam Reactor

    International Nuclear Information System (INIS)

    Holden, N.E.; Hu, J.P.; Reciniello, R.N.

    1998-02-01

    The HFBR is a heavy water, D 2 O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of 235 U. The core is 53 cm high and 48 cm in diameter and has an active volume of 97 liters. The HFBR, which was designed to operate at forty mega-watts, 40 NW, was upgraded to operate at 60 NW. Since 1991, it has operated at 30 MW. In a normal 30 MW operating cycle the HFBR operates 24 hours a day for thirty days, with a six to fourteen day shutdown period for refueling and maintenance work. While most reactors attempts to minimize the escape of neutrons from the core, the HFBR's D 2 O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9. The HFBR neutron dosimetry effort described here compares measured and calculated energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles

  5. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  6. TORT application in reactor pressure vessel neutron flux calculations

    International Nuclear Information System (INIS)

    Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

    1994-01-01

    The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

  7. Equipment for thermal neutron flux measurements in reactor R2

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Nilsson, T; Claeson, S

    1960-04-15

    For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

  8. Behaviour of aged and new flux detectors in Darlington reactors

    Energy Technology Data Exchange (ETDEWEB)

    Banica, C.; Foster, M., E-mail: Constantin.Banica@OPG.com [Ontario Power Generation, Darlington Nuclear, Bowmanville, Ontario (Canada)

    2013-07-01

    In-core neutron flux detectors are used for protective and safety functions in the Darlington NGS 'A' CANDU reactors. This paper presents new observations regarding the aging of flux detectors, including response to fuelling, response to unit shutdown and indicators of detector noise. Comparisons of detector signals before and after replacement confirm previous assumptions about aging effects. (author)

  9. Flux measurement in ZBR at the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Dauke, M.

    2005-01-01

    The determination of the neutron flux in the TRIGA-2-Vienna reactor was the objective of this research. The theory of the method (4π-β detectors) is presented as well as the determination of the maximum flux, gold-cadmium differential measurement, cobalt-wire measurement, finally a comparison of all results was made and interpreted. (nevyjel)

  10. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    Carew, J.F.; Neogy, P.

    1983-01-01

    An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

  11. Boiling water reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

    1975-01-01

    Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

  12. Experiments on Critical Heat Flux for CAREM -25 Reactor

    International Nuclear Information System (INIS)

    Mazufri, C.M

    2000-01-01

    The prediction of critical heat flux (CHF) in rod bundles of light water reactors is basically performed with the aid of empirical correlations derived from experimental data.Many CHF correlations have been proposed and are widely used in the analysis of the thermal margin during normal operation, transient, and accident conditions.Correlations found in the open literature are not sufficiently verified for the thermal hydraulic conditions that appear in the CAREM core under normal operation: high pressure, low flow, and low qualities.To compensate this deficiency, an experimental investigation on CHF in such thermal-hydraulic conditions was carried out.The experiments have been performed in the Institute of Physics and Power Engineering of Russian Federation.A short description of facilities, details of the experimental program and some preliminary results obtained are presented in this work

  13. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1970-04-01

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  14. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  15. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    International Nuclear Information System (INIS)

    Cartier, J.; Casoli, P.; Chappert, F.

    2013-01-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  16. Pursuing nuclear energy with no nuclear contamination - from neutron flux reactor to deuteron flux reactor

    International Nuclear Information System (INIS)

    Li, X. Z.; Wei, Q. M.; Liu, B.; Zhu, X. G.; Ren, S. L.

    2007-01-01

    Pursuing nuclear energy with no nuclear contamination has been a long endeavor since the first fission reactor in 1942. Four major concepts have been the key issues: i.e. resonance, negative feed back, self-sustaining, nuclear radiation. When nuclear energy was just discovered in laboratory, the key issue was to enlarge it from the micro-scale to the macro-scale. Slowing-down the neutrons was the key issue to enhance the fission cross-section in order to build-up the neutron flux through the chain-reactions using resonance between neutron and fissile materials. Once the chain-reaction was realized, the negative feed-back was the key issue to keep the neutron flux at the allowable level. The negative reaction coefficient was introduced by the thermal expansion, and the resonant absorption in cadmium or boron was used to have a self-sustaining fission reactor with neutron flux. Then the strong neutron flux became the origin of all nuclear contamination, and a heavy shielding limits the application of the nuclear energy. The fusion approach to nuclear energy was much longer; nevertheless, it evolved with the similar issues. The resonance between deuteron and triton was resorted to enlarge the fusion cross section in order to keep a self-sustaining hot plasma. However, the 14 MeV neutron emission became the origin of all nuclear contamination again. Deuteron plus helium-3 fusion reaction was proposed to avoid neutron emission although there are two more difficulties: the helium-3 is supposed to be carried back from the moon; and much more higher temperature plasma has to be confined while 50 years needed to realized the deuteron-triton plasma already. Even if deuteron plus helium-3 fusion plasma might be realized in a much higher temperature plasma, we still have the neutron emission from the deuteron-deuteron fusion reaction in the deuteron plus helium-3 fusion plasma. Polarized deuteron-deuteron fusion reaction was proposed early in 1980's to select the neutron

  17. Annual report 1989 operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1989-01-01

    In 1989 the operation of the High Flux Reactor Petten was carried out as planned. The availability was more than 100% of scheduled operating time. The average occupation of the reactor by experimental devices was 72% of the practical occupation limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons and for radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  18. Annual report 1990. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1990-01-01

    In 1990 the operation of the High Flux Reactor was carried out as planned. The availability was 96% of scheduled operating time. The average utilization of the reactor was 71% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  19. Annual Report 1991. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1992-01-01

    In 1991 the operation of the High Flux Reactor was carried out as planned. The availability was more than 100% of scheduled operating time. The average utilization of the reactor was 69% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. Development activities addressed upgrading of irradiation devices, neutron capture therapy, neutron radiography and neutron transmutation doping of silicon. General activities in support of running irradiation programmes progressed in the normal way

  20. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  1. Next generation light water reactors

    International Nuclear Information System (INIS)

    Omoto, Akira

    1992-01-01

    In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

  2. Advanced light-water reactors

    International Nuclear Information System (INIS)

    Golay, M.W.; Todreas, N.E.

    1990-01-01

    Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

  3. Safety and quality management at the high flux reactor Petten

    International Nuclear Information System (INIS)

    Zurita, A.; Ahlf, J.

    1995-01-01

    The High Flux Reactor (HFR) is one high power multi-purpose materials testing research reactor of the tank-in-pool type, cooled and moderated by light-water. It is operated at 45 MW at a prescribed schedule of 11 cycles per year, each comprising 25 operation days and three shut-down days. Since the licence for the operation of HFR was granted in 1962, a total of 14 amendments to the original licence have been made following different modifications in the installations. In the meantime, international nuclear standards were developed, especially in the framework of the NUSS programme of the IAEA, which were adopted by the Dutch Licensing Authorities. In order to implement the new standards, the situation at the HFR was comprehensively reviewed in the course of an audit performed by the Dutch Licensing Authorities in 1988. This also resulted in formulating the task of setting-up an 'HFR - Integral Quality Assurance Handbook' (HFR-IQAD) involving both organizations JRCIAM and ECN, which had the unique framework and basic guideline to assure the safe and efficient operation and exploitation of the HFR and to promote safety and quality in all aspects of HFR related activities. The assurance of safe and efficient operation and exploitation of the HFR is condensed together under the concepts of safety and quality of services and is achieved through the safety and quality management. (orig.)

  4. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  5. Chemistry in water reactors

    International Nuclear Information System (INIS)

    Hermansson, H.P.; Norring, K.

    1994-01-01

    The international conference Chemistry in Water Reactors was arranged in Nice 24-27/04/1994 by the French Nuclear Energy Society. Examples of technical program areas were primary chemistry, operational experience, fundamental studies and new technology. Furthermore there were sessions about radiation field build-up, hydrogen chemistry, electro-chemistry, condensate polishing, decontamination and chemical cleaning. The conference gave the impression that there are some areas that are going to be more important than others during the next few years to come. Cladding integrity: Professor Ishigure from Japan emphasized that cladding integrity is a subject of great concern, especially with respect to waterside corrosion, deposition and release of crud. Chemistry control: The control of the iron/nickel concentration quotient seems to be not as important as previously considered. The future operation of a nuclear power plant is going to require a better control of the water chemistry than achievable today. One example of this is solubility control via regulation in BWR. Trends in USA: means an increasing use of hydrogen, minimization of SCC/IASCC, minimization of radiation fields by thorough chemistry control, guarding fuel integrity by minimization of cladding corrosion and minimization of flow assisted corrosion. Stellite replacement: The search for replacement materials will continue. Secondary side crevice chemistry: Modeling and practical studies are required to increase knowledge about the crevice chemistry and how it develops under plant operation conditions. Inhibitors: Inhibitors for IGSCC and IGA as well for the primary- (zinc) as for the secondary side (Ti) should be studied. The effects and mode of operation of the inhibitors should be documented. Chemical cleaning: of heat transfer surfaces will be an important subject. Prophylactic cleaning at regular intervals could be one mode of operation

  6. Flux distribution measurements in the Bruce A unit 1 reactor

    International Nuclear Information System (INIS)

    Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

    1977-07-01

    Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

  7. Flux-limited diffusion coefficients in reactor physics applications

    International Nuclear Information System (INIS)

    Pounders, J.; Rahnema, F.; Szilard, R.

    2007-01-01

    Flux-limited diffusion theory has been successfully applied to problems in radiative transfer and radiation hydrodynamics, but its relevance to reactor physics has not yet been explored. The current investigation compares the performance of a flux-limited diffusion coefficient against the traditionally defined transport cross section. A one-dimensional BWR benchmark problem is examined at both the assembly and full-core level with varying degrees of heterogeneity. (authors)

  8. The heavy water reactors

    International Nuclear Information System (INIS)

    Brudermueller, G.

    1976-01-01

    This is a survey of the development so far of this reactor line which is in operation all over the world in various types (e.g. BHWR, PHWR). MZFR and the CANDU-type reactors are discussed in more detail. (UA) [de

  9. Nuclear reactor in deep water

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

  10. 1984 Operation of the high flux reactor

    International Nuclear Information System (INIS)

    1985-01-01

    The programme resources in 1984 were largely devoted to the replacement of the old reactor vessel and its peripheral equipment. The original vessel had been in operation for more than 20 years and doubts had arisen about the condition of the aluminium tank after so long an exposure to neutrons. The operation, which had never been attempted before on a reactor of that size and complexity was planned and prepared over a number of years to take advantage of the occasion to provide a much improved vessel, incorporating the latest design features. The plant was shut down at the end of November 1983 and the 14 months operation began with a short cooling-off period for decay of short lived radioactivity followed by removal of the old tank and its dissection into pieces convenient for consolidation and storage as radioactive waste. After decontamination of the shielding pool, the new vessel and neutron beam tubes were installed and the reactor was recommissioned. Routine 45 MW operation was resumed on 14 February 1985 and has been uneventful since then

  11. Critical heat flux correlation analysis for PWR reactors with low mass flow

    International Nuclear Information System (INIS)

    Carajilescov, Pedro

    1996-01-01

    The major limit in the thermalhydraulic design of water cooled reactors consists in the occurrence of critical heat flux, which is verified by correlation of large range of validity. In the present work, the major design correlations were analyzed, through comparisons with experimental data, for utilization in PWR with low mass flux in the core. The results show that the EPRI correlation, with modifications, gives conservative results, from the safety point of view, with lower data spreading, being the most indicated for the reactor thermal design. (author)

  12. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  13. Preliminary study of a flux converter for experimental reactor

    International Nuclear Information System (INIS)

    Malouch, M.F.

    1998-01-01

    The purpose of this project is to define the characteristics of a flux converter dedicated to increase the fast neutron flux in irradiation devices placed in the core of Osiris experimental reactor. This preliminary work has dealt with the neutronic and thermal-hydraulic aspects of this problem. The synthesis of the results produced by the codes APOLLO2, DAIXY, MERCURE5.3 and FLICA-3M shows that a cylindrical converter equipped with 5 fissile rings can enhance the fast flux by a 35% factor in an experimental device set in its center. (A.C.)

  14. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  15. Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

    International Nuclear Information System (INIS)

    Rana, Y.S.; Mishra, Abhishek; Singh, Tej

    2016-06-01

    Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

  16. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  17. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  18. 1982 Annual status report: operation of the high flux reactor

    International Nuclear Information System (INIS)

    1983-01-01

    The high flux materials testing reactor has been operated in 1982 within a few percent of the pre-set schedule, attaining 73% overall availability. Its utilization reached another record figure in 20 years: 81% without, 92% with, the low enrichment test elements irradiated during the year

  19. Transient neutrons flux behaviour in a spherical reactor core

    International Nuclear Information System (INIS)

    Souza, A.W.A. de.

    1978-11-01

    This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

  20. Water simulation of sodium reactors

    International Nuclear Information System (INIS)

    Grewal, S.S.; Gluekler, E.L.

    1981-01-01

    The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

  1. One procedure for determination of the neutron flux in the nuclear reactor fuel

    International Nuclear Information System (INIS)

    Bulovic, V.; Krtil, J.; Maksimovic, Z.; Martinc, R.

    1979-09-01

    Possibility of determination of the neutron flux in the fuel of a heavy water reactor has been examined. In determination of the flux an iterative procedure was used to compare calculated and measured contents of several fission products. The former contents were determined by calculation of the burning process balance and the latter by non-destructive gamma-spectrometric analysis of fuel. The obtained results prove the possibility of such determination of not only the average value of the flux but also of the change of its intensity during utilization of fuel (author) [sr

  2. Device for detecting neutron flux in nuclear reactor. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Bessho, Y; Nishizawa, Y

    1976-07-30

    The object of the invention is to ensure accuracy in the operation of the nuclear reactor by reducing the difference that results between the readings of a Traversing Incore Probe (TIP) and a Local Power Range Monitor (LPRM) when the neutron flux distribution undergoes a change. In an apparatus for detecting neutrons in a nuclear reactor, an LPRM sensor comprising a layer containing a substance capable of nuclear fission, a section filled with argon gas and a collector is constructed so as to surround a TIP within a TIP guide tube at the height of the reactor axis. In this way, the LPRM detects the average value of neutron distribution in the region surrounding the TIP, so that no great difference between the readings of both the sensors is produced even if the neutron flux distribution is changed.

  3. Instrumentation for the advanced high-flux reactor workshop: proceedings

    International Nuclear Information System (INIS)

    Moon, R.M.

    1984-01-01

    The purpose of the Workshop on Instrumentation for the Advanced High-Flux Reactor, held on May 30, 1984, at the Oak Ridge National Laborattory, was two-fold: to announce to the scientific community that ORNL has begun a serious effort to design and construct the world's best research reactor, and to solicit help from the scientific community in planning the experimental facilities for this reactor. There were 93 participants at the workshop. We are grateful to the visiting scientists for their enthusiasm and interest in the reactor project. Our goal is to produce a reactor with a peak thermal flux in a large D 2 O reflector of 5 x 10 15 n/cm 2 s. This would allow the installation of unsurpassed facilities for neutron beam research. At the same time, the design will provide facilities for isotope production and materials irradiation which are significantly improved over those now available at ORNL. This workshop focussed on neutron beam facilities; the input from the isotope and materials irradiation communities will be solicited separately. The reactor project enjoys the full support of ORNL management; the present activities are financed by a grant of $663,000 from the Director's R and D Fund. However, we realize that the success of the project, both in realization and in use of the reactor, depends on the support and imagination of a broad segment of the scientific community. This is more a national project than an ORNL project. The reactor would be operated as a national user facility, open to any research proposal with high scientific merit. It is therefore important that we maintain a continuing dialogue with outside scientists who will be the eventual users of the reactor and the neutron beam facilities. The workshop was the first step in establishing this dialogue; we anticipate further workshops as the project continues

  4. Very high flux steady state reactor and accelerator based sources

    International Nuclear Information System (INIS)

    Ludewig, H.; Todosow, M.; Simos, N.; Shapiro, S.; Hastings, J.

    2004-01-01

    With the number of steady state neutron sources in the US declining (including the demise of the Bnl HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from 10 14 n/cm 2 *s to 10 15 n/cm 2 *s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (about 350 MW) with a projected undisturbed thermal flux of 7*10 15 n/cm 2 *s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 10 16 n/cm 2 *s. The first is a reactor based concept operating at high power density (10 MW/l - 15 MW/l) and a total power of 100 MW - 250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV - 1.5 GeV proton Linac with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study, the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated. (authors)

  5. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    Drevon, G.

    1983-01-01

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

  6. High flux reactor evolutions and improvements

    International Nuclear Information System (INIS)

    Guyon, H.

    2005-01-01

    Following the changes over the years in experimental and safety requirements at the ILL a great deal of work has been carried out on the installations: - In 1985, a new cold source was installed, allowing the production of ultra-cold neutrons via a vertical channel. - From 1991 to 1995 the reactor block was replaced, allowing us to perform reactivity calculations and determine other neutronic values. - In 2003, a new hot source was installed with three beam tubes viewing it; the new system is now operating very efficiently. - This year a major beam tube is to be replaced with a new zircaloy tube. - And finally, from 2003 to 2006, the facility is being upgraded significantly to withstand newly-defined safe-shutdown earthquakes. In parallel, developments are on-going on the efficiency of the instruments and the neutron guides under the Millennium Programme. These will result in overall gains in data collection of over a factor of 10. As the ILL's international convention has been extended to the end of 2013 the Institute is therefore now well-set to maintain its position as a centre of excellence in the scientific use of slow neutrons for the twenty years to come. (author)

  7. Devices for irradiation of materials in the fast neutron flux at the RA heavy water reactor in Vinca; Uredjaji za ozracivanje materijala u fluksu brzih neutrona na teskovodnom reaktoru RA u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-07-01

    Full text: This paper covers the concept, design description and problems of constructing special experimental devices at the RA reactor in Vinca. These devices were used for irradiation of construction materials (graphite, Mg and Al oxides, steel and zircaloy) in the integral fast neutron flux of 2 x 10{sup 20} n/cm{sup 2}. Temperature of samples cooled by heavy water circulating through the hollow uranium fuel elements was measured by thermocouples. Construction of the device, preparation of samples, related flux measurements as well as irradiation were done in cooperation with the nuclear center in Saclay, France. Brief review of the experiences gained in using these experimental spaces at the RA reactor as well as problems of transporting the irradiated capsules from Vinca to France is included. [Serbo-Croat] Pun tekst: Rad obuhvata koncepciju, tehnicki opis i probleme realizacije specijalnih eksperimentalnih uredjaja u reaktoru RA u Vinci, u kojima je tokom 1962. godine izvrseno, na temperaturama ispod 100 deg C, ozracivanje konstrukcionih materijala (grafita, Mg i Al-oksida, celika i cirkaloja) u integralnom fluksu brzih neutrona vrednosti 2 x 10{sup 20} n/cm{sup 2}. Temperatura uzoraka hladjenih teskom vodom, koja protice kroz suplje uranske elemente, merena je termoparovima. Realizacija uredjaja, priprema uzoraka, odgovarajuca merenja fluksa kao i sama ozracivanja vrsena su u saradnji sa Nuklearnim centrom u Saclay-u, Francuska. Dat je kraci pregled iskustava iz eksploatacije ovih novih eksperimentalnih prostora u reaktoru RA, kao i neki problemi vezani za transport aktivnih kapsula sa uzorcima od Vince do Saclay-a (author)

  8. The High Flux Beam Reactor at Brookhaven National Laboratory

    International Nuclear Information System (INIS)

    Shapiro, S.M.

    1994-01-01

    Brookhaven National Laboratory's High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want 'more'. In the mid-50's the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments

  9. Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

    Science.gov (United States)

    Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

    2012-08-01

    The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

  10. Heat exchangers in heavy water reactor systems

    International Nuclear Information System (INIS)

    Mehta, S.K.

    1988-01-01

    Important features of some major heat exchange components of pressurized heavy water reactors and DHRUVA research reactor are presented. Design considerations and nuclear service classifications are discussed

  11. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cartier, J. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France); Casoli, P. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, Valduc, F-21120 Is sur Tille (France); Chappert, F. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France)

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  12. PODESY program for flux mapping of CNA II reactor:

    International Nuclear Information System (INIS)

    Ribeiro Guevara, Sergio

    1988-01-01

    The PODESY program, developed by KWU, calculates the spatial flux distribution of CNA II reactor through a three-dimensional expansion of 90 incore detector measurements. The calculation is made in three steps: a) short-term calculation which considers the control rod positions and it has to be done each time the flux mapping is calculated; b) medium-term calculation which includes local burn-up dependent calculation made by diffusion methods in macro-cell configurations (seven channels in hexagonal distribution), and c) long-term calculation, or macroscopic flux determination, that is a fitting and expansion of measured fluxes, previously corrected by local effects, using the eigen functions of the modified diffusion equation. The paper outlines development of step (c) of the calculation. The incore detectors have been located in the central zone of the core. In order to obtain low errors in the expansion procedure it is necessary to include additional points, whose flux values are assumed to be equivalent to detector measurements. These flux values are calculated with detector measurements and a spatial flux distribution calculated by a PUMA code. This PUMA calculation employs a smooth burn-up distribution (local burn-up variations are considered in step (b) of the whole calculation) representing the state of core evolution at the calculation time. The core evolution referred to ends when the equilibrium core condition is reached. Additionally, a calculation method to be employed in the plant in case of incore detector failures, is proposed. (Author) [es

  13. Pressurizer pump reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

  14. Instrument lance for boiling water reactors

    International Nuclear Information System (INIS)

    Proell, N.; Bertz, S.; Graebener, K.H.

    1980-01-01

    The instrument lance contains in the lance cover pipe a thimble as part of the drive chamber system. Other thimbles serve to carry neutron detectors. Detectors can be exchanged without opening the reactor pressure vessel and without removing the fuel elements. Furthermore the detector exchange is independent from the fuel element cycle. The measurement lance passes from the bottom of the pressure vessel over the total hight of the core in the water ducts between the fuel elements and can thus determine the neutron flux distribution. (DG) [de

  15. Decay ratio estimation in pressurized water reactor

    International Nuclear Information System (INIS)

    Por, G.; Runkel, J.

    1990-11-01

    The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

  16. Thermal flux flattering and increase of reactor output

    Energy Technology Data Exchange (ETDEWEB)

    Horowitz, J; Bussac, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1959-07-01

    It is worthwhile, when building power reactors, to have excess reactivity in order to increase rating by fitting closely together the heat sources and the cooling possibilities. The power per unit volume of a graphite reactor can then be increased, given the power of the most heavily loaded channel. The solutions adopted for G.1, G.2, and E.D.F.1 are described here, and also the improvements based on the actual neutron flux flattening, the introduction of several zones for the coolant, the variation of uranium rod and coolant channel diameters according to their location, and finally the change in lattice pitch. The perturbation of neutron flux due to variation of mean absorption in the lattice is also discussed. (author)

  17. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  18. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    Energy Technology Data Exchange (ETDEWEB)

    Diaz Rizo, O; Alvarez, I; Herrera, E; Lima, L; Tores, J [Secretaria Ejecutiva para Asuntos Nucleares, Holguin (Cuba). Delegacion Territorial; Manso, M V [Centro de Isotopos, La Habana (Cuba); Lopez, M C [Instituto Nacional de Investigaciones Nucleares, Mexico City (Mexico); Ixquiac, M [Universidad de San Carlos de Guatemala, Guatemala City (Guatemala)

    1997-12-31

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K{sub o} neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott`s formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented.

  19. Annual progress report 1988, operation of the high flux reactor

    International Nuclear Information System (INIS)

    1989-01-01

    In 1988 the High Flux Reactor Petten was routinely operated without any unforeseen event. The availability was 99% of scheduled operation. Utilization of the irradiation positions amounted to 80% of the practical occupation limit. The exploitation pattern comprised nuclear energy deployment, fundamental research with neutrons, and radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  20. Component and system simulation models for High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs

  1. Thorium in heavy water reactors

    International Nuclear Information System (INIS)

    Andersson, G.

    1984-12-01

    Advanced heavy water reactors can provide energy on a global scale beyond the foreseeable future. Their economic and safety features are promising: 1. The theoretical feasibility of the Self Sufficient Equilibrium Thorium (SSET) concept is confirmed by new calculations. Calculations show that the adjuster rod geometry used in natural uranium CANDU reactors is adequate also for SSET if the absorption in the rods is graded. 2. New fuel bundle designs can permit substantially higher power output from a CANDU reactor. The capital cost for fuel, heavy water and mechanical equipment can thereby be greatly reduced. Progress is possible with the traditional fuel material oxide, but the use of thorium metal gives much larger effects. 3. A promising long range possibility is to use pressure tanks instead of pressure tubes. Heat removal from the core is facilitated. Negative temperature and void coefficients provide inherent safety features. Refuelling under power is no longer needed if control by moderator displacement is used. Reduced quality demand on the fuel permits lower fuel costs. The neutron economy is improved by the absence of pressure and clandria tubes and also by the use of radial and axial blankets. A modular seed blanket design can reduce the Pa losses. The experience from construction of tank designs is good e.g. AAgesta, Attucha. It is now also possible to utilize technology from LWR reactors and the implementation of advanced heavy water reactors would thus be easier than HTR or LMFBR systems. (Author)

  2. Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

    International Nuclear Information System (INIS)

    Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

    2011-01-01

    Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

  3. Operation of the High Flux Reactor. Annual report 1985

    International Nuclear Information System (INIS)

    1985-01-01

    This year was characterized by the end of a major rebuilding of the installation during which the reactor vessel and its peripheral components were replaced by new and redesigned equipment. Both operational safety and experimental use were largely improved by the replacement. The reactor went back to routine operation on February 14, 1985, and has been operating without problem since then. All performance parameters were met. Other upgrading actions started during the year concerned new heat exchangers and improvements to the reactor building complex. The experimental load of the High Flux Reactor reached a satisfactory level with an average of 57%. New developments aimed at future safety related irradiation tests and at novel applications of neutrons from the horizontal beam tubes. A unique remote encapsulation hot cell facility became available adding new possibilities for fast breeder fuel testing and for intermediate specimen examination. The HFR Programme hosted an international meeting on development and use of reduced enrichment fuel for research reactors. All aspects of core physics, manufacture technology, and licensing of novel, proliferation-free, research reactor fuel were debated

  4. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

    2011-01-01

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

  5. Fuel management at the Petten high flux reactor

    International Nuclear Information System (INIS)

    Thijssen, P.J.M.

    1999-01-01

    Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high 235 U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing facilities can be used for many different types of application. (author)

  6. Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2007-01-01

    Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

  7. Hydrogen water chemistry for boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

    1985-01-01

    Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

  8. Modern Estimates of Global Water Cycle Fluxes

    Science.gov (United States)

    Rodell, M.; Beaudoing, H. K.; L'Ecuyer, T. S.; Olson, W. S.

    2014-12-01

    The goal of the first phase of the NASA Energy and Water Cycle Study (NEWS) Water and Energy Cycle Climatology project was to develop "state of the global water cycle" and "state of the global energy cycle" assessments based on data from modern ground and space based observing systems and data integrating models. Here we describe results of the water cycle assessment, including mean annual and monthly fluxes over continents and ocean basins during the first decade of the millennium. To the extent possible, the water flux estimates are based on (1) satellite measurements and (2) data-integrating models. A careful accounting of uncertainty in each flux was applied within a routine that enforced multiple water and energy budget constraints simultaneously in a variational framework, in order to produce objectively-determined, optimized estimates. Simultaneous closure of the water and energy budgets caused the ocean evaporation and precipitation terms to increase by about 10% and 5% relative to the original estimates, mainly because the energy budget required turbulent heat fluxes to be substantially larger in order to balance net radiation. In the majority of cases, the observed annual, surface and atmospheric water budgets over the continents and oceans close with much less than 10% residual. Observed residuals and optimized uncertainty estimates are considerably larger for monthly surface and atmospheric water budget closure, often nearing or exceeding 20% in North America, Eurasia, Australia and neighboring islands, and the Arctic and South Atlantic Oceans. The residuals in South America and Africa tend to be smaller, possibly because cold land processes are a non-issue. Fluxes are poorly observed over the Arctic Ocean, certain seas, Antarctica, and the Australasian and Indonesian Islands, leading to reliance on atmospheric analysis estimates. Other details of the study and future directions will be discussed.

  9. Contribution to the qualification of calculation methods of reactivity and of flux and power distributions in nuclear pressurized water reactor cores

    International Nuclear Information System (INIS)

    Abit, K.

    1984-01-01

    The last stage of the creation computer methods and calculations consists of verifying the running and qualifying the results obtained. The work of the present thesis consisted of improving a coupling method between radial and axial phenomena in a PWR core, refering to three-dimensional calculations, while ensuring a perfect coherence between the programmed physical models. The calculation results have been compared to measurements of reactivity and of flux distributions realized during start-up tests. Thus, the methods have been applied to the calculation of the evolution of a burnable poison (gadolinium) in view of operation in long campaign. 13 refs [fr

  10. Terrestrial water fluxes dominated by transpiration.

    Science.gov (United States)

    Jasechko, Scott; Sharp, Zachary D; Gibson, John J; Birks, S Jean; Yi, Yi; Fawcett, Peter J

    2013-04-18

    Renewable fresh water over continents has input from precipitation and losses to the atmosphere through evaporation and transpiration. Global-scale estimates of transpiration from climate models are poorly constrained owing to large uncertainties in stomatal conductance and the lack of catchment-scale measurements required for model calibration, resulting in a range of predictions spanning 20 to 65 per cent of total terrestrial evapotranspiration (14,000 to 41,000 km(3) per year) (refs 1, 2, 3, 4, 5). Here we use the distinct isotope effects of transpiration and evaporation to show that transpiration is by far the largest water flux from Earth's continents, representing 80 to 90 per cent of terrestrial evapotranspiration. On the basis of our analysis of a global data set of large lakes and rivers, we conclude that transpiration recycles 62,000 ± 8,000 km(3) of water per year to the atmosphere, using half of all solar energy absorbed by land surfaces in the process. We also calculate CO2 uptake by terrestrial vegetation by connecting transpiration losses to carbon assimilation using water-use efficiency ratios of plants, and show the global gross primary productivity to be 129 ± 32 gigatonnes of carbon per year, which agrees, within the uncertainty, with previous estimates. The dominance of transpiration water fluxes in continental evapotranspiration suggests that, from the point of view of water resource forecasting, climate model development should prioritize improvements in simulations of biological fluxes rather than physical (evaporation) fluxes.

  11. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  12. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  13. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  14. Utilization of cold neutron beams at intermediate flux reactors

    International Nuclear Information System (INIS)

    Clark, D.D.

    1992-01-01

    With the advent of cold neutron beam (CNB) facilities at U.S. reactors [National Institute of Standards and Technology (NIST) in 1991; Cornell University and the University of Texas at Austin, anticipated in 1992], it is appropriate to reexamine the types of research for which they are likely to be best suited or uniquely suited. With the exception of a small-angle neutron scattering facility at Brookhaven National Laboratory, there has been no prior experience in the United States with such beams, but they have been extensively used at European reactors where cold neutron sources and neutron guides were developed some years age. This paper does not discuss specialized cases such as ultracold neutrons or very high flux facilities such as the Institute Laue-Langevin ractor and the proposed advanced neutron source. Instead, it concentrates on potential utilization of CNBs at intermediate-flux reactors such as at Cornell and Texas, i.e., in the 1-MW range and operated <24 h a day

  15. Absolute measurement of neutron fluxes inside the reactor core

    International Nuclear Information System (INIS)

    Ajdacic, S. V.

    1964-10-01

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li 6 -semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li 6 -semiconductor spectrometer with plane geometry is given. A new type of Li 6 -semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li 6 -spectrometer made (author)

  16. Absolute measurement of neutron fluxes inside the reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Ajdacic, S V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-10-15

    The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li{sup 6}-semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li{sup 6}-semiconductor spectrometer with plane geometry is given. A new type of Li{sup 6}-semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li{sup 6}-spectrometer made (author)

  17. The safety of light water reactors

    International Nuclear Information System (INIS)

    Pershagen, B.

    1986-04-01

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  18. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

  19. Terrestrial water fluxes dominated by transpiration: Comment

    Science.gov (United States)

    Daniel R. Schlaepfer; Brent E. Ewers; Bryan N. Shuman; David G. Williams; John M. Frank; William J. Massman; William K. Lauenroth

    2014-01-01

    The fraction of evapotranspiration (ET) attributed to plant transpiration (T) is an important source of uncertainty in terrestrial water fluxes and land surface modeling (Lawrence et al. 2007, Miralles et al. 2011). Jasechko et al. (2013) used stable oxygen and hydrogen isotope ratios from 73 large lakes to investigate the relative roles of evaporation (E) and T in ET...

  20. Distributed plant simulator with two-phase flow analysis code using drift-flux non-equilibrium model for pressurized water reactors

    International Nuclear Information System (INIS)

    Yamamoto, Takaya; Kitamura, Masashi; Ohi, Tadashi; Akagi, Katsumi

    1999-01-01

    As advanced monitoring and controlling systems, such as the advanced main control console and the operator support system have been developed, real-time simulators' simulation accuracy must be improved and simulation limits must be extended. Therefore the authors have developed a distributed simulation system to achieve high processing performance using low cost hardware. Moreover, the authors have developed a thermal-hydraulic computer code, using drift-flux non-equilibrium model, which can realize a high precision two-phase flow analysis, which is considered to have the same prediction capability as two-fluid models, while achieving high speed and stability for real-time simulators. The distributed plant simulator for PWR plants was realized as a result. The distributed simulator consists of multi-processors connected to each other by an optical fiber network. Controlling software for synchronized scheduling and memory transfer was also developed. The simulation results of the four loop PWR simulator are compared with experimental data and real plant data; the agreement is satisfactory for a plant simulator. The simulation speed is also satisfactory being twice as fast as real-time. (author)

  1. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  2. Testing of research reactor fuel in the high flux reactor (Petten)

    International Nuclear Information System (INIS)

    Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

    1999-01-01

    The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

  3. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    Date, V.G.; Kulkarni, P.G.

    2000-01-01

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  4. Lessons learned form high-flux isotope reactor restart efforts

    International Nuclear Information System (INIS)

    Dahl, T.L.

    1989-01-01

    When the high-flux isotope reactor's (HFIR's) pressure vessel irradiation surveillance specimens were examined in December 1986, unexpected embrittlement was found. The resulting investigation disclosed widespread deficiencies in quality assurance and management practices. On March 24, 1987, the US Department of Energy (DOE) mandated a shutdown of all five Oak Ridge National Laboratory (ORNL) research reactors. Since the beginning of 1987, 18 different formal review groups have evaluated the management and operations of the HFIR. The root cause of the identified deficiencies in the HFIR program was defined as a lack of rigor in management practices and complacency built on twenty years of trouble-free operation. A number of lessons can be learned from the HFIR experience. Particular insight can be gained by comparing the HFIR organization prior to the shutdown with the organization that exists today. Key elements in such a comparison include staffing, funding, discipline, and formality in operations, maintenance, and management

  5. Status in 1998 of the high flux reactor fuel cycle

    International Nuclear Information System (INIS)

    Guidez, J.; Gevers, A.; Wijtsma, F.J.; Thijssen, P.M.J.

    1998-01-01

    The High Flux Reactor located at Petten (The Netherlands), is owned by the European Commission and is operated under contract by ECN (Netherlands Energy Research Foundation). This plant is in operation since 1962 using HEU enriched at 90%. Conversion studies were conducted several years ago with the hypothesis of a global conversion of the entire core. The results of these studies have shown a costly operation with a dramatic decrease of the thermal flux which is necessary for the medical use of the plant (Molybdene 99 production). Some tests with low enriched elements were also conducted with several companies, several geometrical configurations and several enrichments. They are described in this paper. Explanations are also given on future possibilities for new fuel testing. (author)

  6. High flux-fluence measurements in fast reactors

    International Nuclear Information System (INIS)

    Lippincott, E.P.; Ulseth, J.A.

    1977-01-01

    Characterization of irradiation environments for fuels and materials tests in fast reactors requires determination of the neutron flux integrated over times as long as several years. An accurate integration requires, therefore, passive dosimetry monitors with long half-life or stable products which can be conveniently measured. In addition, burn-up, burn-in, and burn-out effects must be considered in high flux situations and use of minimum quantities of dosimeter materials is often desirable. These conditions force the use of dosimeter and dosimeter container designs, measured products, and techniques that are different from those that are used in critical facilities and other well-characterized benchmark fields. Recent measurements in EBR-II indicate that high-accuracy results can be attained and that tie-backs to benchmark field technique calibrations can be accomplished

  7. Study of the neutron flux distribution in acylindrical reactor

    Directory of Open Access Journals (Sweden)

    A. Vidal-Ferràndiz

    2017-08-01

    Full Text Available In the Energy Engineering Degree of the Universitat Politècnica de València, the students attend to the Nuclear Technology course, in which the basic knowledge of this technology is presented. A main objective of this technology is to obtain neutron population distribution inside a reactor core, in order to maintain the fission reaction chain. As this activity cannot be experimentally developed, mathematical modelling is of great importance to achieve such objective.  One of the computer laboratories proposed consists in the neutron flux determination analytically and numerically in a cylindrical geometry. The analytical solution makes use of the Bessel functions and is a good example of their applications. Alternatively, a numerical solution based on finite differences is used to obtain an approximate solution of the neutron flux. In this work, different discretizations of the cylindrical geometry are implemented and their results are compared.

  8. Seismic upgrading of the Brookhaven High Flux Beam Research Reactor

    International Nuclear Information System (INIS)

    Subudhi, M.

    1985-01-01

    In recent years the High Flux Beam Research (HFBR) reactor facility at Brookhaven National Laboratory (BNL) was upgraded from 40 to 50 MW power level. The reactor plant was built in the early sixties to the seismic design requirements of the period, using the static load approach. While the plant power level was upgraded, the seismic design was also improved according to current design criteria. This included the development of new floor response spectra for the facility and an overall seismic analysis of those systems important to the safe shutdown of the reactor. Items included in the reanalysis are the containment building with its internal structure, the piping systems, tanks, equipment, and heat exchangers. This paper describes the procedure utilized in developing the floor response spectra for the existing facility. Also included in the paper are the findings and recommendations, based on the seismic analysis, regarding the seismic adequacy of structural and mechanical systems vital to achieving the safe shutdown of the reactor. 11 references, 4 figures, 1 table

  9. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    1994-03-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  10. Analysis of an accelerator-driven subcritical light water reactor

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

    1997-01-01

    An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

  11. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  12. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L., E-mail: mabs500@gmail.com, E-mail: narain@ufpe.br, E-mail: jairbezerra@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociências. Departamento de Energia Nuclear

    2017-07-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  13. Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

    International Nuclear Information System (INIS)

    Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L.

    2017-01-01

    Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

  14. High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

    Energy Technology Data Exchange (ETDEWEB)

    Pudelek, R. E.; Gilbert, W. C.

    2002-02-26

    This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste

  15. Feed water control device in a reactor

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1984-01-01

    Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

  16. Water treatment process for nuclear reactors

    International Nuclear Information System (INIS)

    Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

    1992-01-01

    Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerates to obtain the optimum quantity of pure water which reached to 15 cubic meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30 %. output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined.5 fig., 3 tab

  17. Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

    International Nuclear Information System (INIS)

    Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-01-01

    The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology

  18. Nuclear reactor and production systems with flux-optical digitizer

    International Nuclear Information System (INIS)

    Luger, P.P.; Nealen, J.P.

    1979-01-01

    Several digital sensing devices are described for use in automated production systems. The first described is for use in the automatic operation of a reactor. This device employs a binant electrometer using a quartz fiber mounted at one end but free to vibrate at the other in an AC field. The fiber oscillates if a charge is placed upon it. An optical slit replaces the ordinary eyepiece reticule scale. With the quartz fiber adjusted so its image is in focus at the optical slit, photoelectric signals are obtained at null charge on the fiber. The quartz fiber is repeatedly charged and allowed to discharge by collecting ions from a source under measurement. Each photoelectric signal causes a digital time reading to be taken. The time readings are used to evaluate the current due to the electric charge. The photoelectric signals, by feedback, also operate the electrometer for continuous intermittent-continuous operation. Basically the current is a current digitizer. Application is made to reactor monitoring and control as well as to other types of production systems. The flux-optical digitizer is a radiometer-like-structure carryig rotating fins that may be coated with fissionable material, such as 235 U for the purpose of neutron flux measurements. The rotating fins are mounted on a shaft that also carries an arm that produces photoelectric signals whenever the arm overlaps an optical slit and thus diminishes light from an auxiliary light flux source incident on the slit. Between successive photoelectric signals, time interval measurements are obtained. This and other sensing devices are fully described for various automated, controlled, production processes

  19. Proposal for a new method of reactor neutron flux distribution determination

    Energy Technology Data Exchange (ETDEWEB)

    Popic, V R [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1964-01-15

    A method, based on the measurements of the activity produced in a medium flowing with variable velocity through a reactor, for the determination of the neutron flux distribution inside a reactor is considered theoretically (author)

  20. Fabrication of control rods for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sease, J.D.

    1998-01-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A

  1. Rebuilding the Brookhaven high flux beam reactor: A feasibility study

    International Nuclear Information System (INIS)

    Brynda, W.J.; Passell, L.; Rorer, D.C.

    1995-01-01

    After nearly thirty years of operation, Brookhaven's High Flux Beam Reactor (HFBR) is still one of the world's premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR's value as a national scientific resource, members of the Laboratory's scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor's research capabilities

  2. Fabrication of control rods for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  3. The epithermal neutron-flux distribution in the reactor RA - Vinca

    International Nuclear Information System (INIS)

    Marinkov, V.; Bikit, I.; Martinc, R.; Veskovic, M.; Slivka, J.; Vaderna, S.

    1987-01-01

    The distribution of the epithermal neutron flux in the reactor RA - Vinca has been measured by means of Zr - activation detectors. In the channel VK-8 non-homogeneous flux distribution was observed (author) [sr

  4. Water treatment process for nuclear reactors

    International Nuclear Information System (INIS)

    Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

    1993-01-01

    Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigates water treatment process for nuclear reactor utilization. Analysis of outwater chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerants to obtain the optimum quantity of pure water which reached to 15 cubic-meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30%. Output water chemistry agrees with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined

  5. Advanced Multiphysics Thermal-Hydraulics Models for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Prashant K [ORNL; Freels, James D [ORNL

    2015-01-01

    Engineering design studies to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from using highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL). This work is part of an effort sponsored by the US Department of Energy (DOE) Reactor Conversion Program. HFIR is a very high flux pressurized light-water-cooled and moderated flux-trap type research reactor. HFIR s current missions are to support neutron scattering experiments, isotope production, and materials irradiation, including neutron activation analysis. Advanced three-dimensional multiphysics models of HFIR fuel were developed in COMSOL software for safety basis (worst case) operating conditions. Several types of physics including multilayer heat conduction, conjugate heat transfer, turbulent flows (RANS model) and structural mechanics were combined and solved for HFIR s inner and outer fuel elements. Alternate design features of the new LEU fuel were evaluated using these multiphysics models. This work led to a new, preliminary reference LEU design that combines a permanent absorber in the lower unfueled region of all of the fuel plates, a burnable absorber in the inner element side plates, and a relocated and reshaped (but still radially contoured) fuel zone. Preliminary results of estimated thermal safety margins are presented. Fuel design studies and model enhancement continue.

  6. Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

    International Nuclear Information System (INIS)

    Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

    1987-01-01

    A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

  7. Quantification of Water Flux in Vesicular Systems.

    Science.gov (United States)

    Hannesschläger, Christof; Barta, Thomas; Siligan, Christine; Horner, Andreas

    2018-06-04

    Water transport across lipid membranes is fundamental to all forms of life and plays a major role in health and disease. However, not only typical water facilitators like aquaporins facilitate water flux, but also transporters, ion channels or receptors represent potent water pathways. The efforts directed towards a mechanistic understanding of water conductivity determinants in transmembrane proteins, the development of water flow inhibitors, and the creation of biomimetic membranes with incorporated membrane proteins or artificial water channels depend on reliable and accurate ways of quantifying water permeabilities P f . A conventional method is to subject vesicles to an osmotic gradient in a stopped-flow device: Fast recordings of scattered light intensity are converted into the time course of vesicle volume change. Even though an analytical solution accurately acquiring P f from scattered light intensities exists, approximations potentially misjudging P f by orders of magnitude are used. By means of computational and experimental data we point out that erroneous results such as that the single channel water permeability p f depends on the osmotic gradient are direct results of such approximations. Finally, we propose an empirical solution of which calculated permeability values closely match those calculated with the analytical solution in the relevant range of parameters.

  8. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    Tas, A.; Teunissen, G.

    1986-01-01

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

  9. Emergency diesel generator reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    A program to apply some of the techniques of reliability engineering to the High Flux Isotope Reactor (HFIR) was started on August 8, 1992. Part of the program was to track the conditional probabilities of the emergency diesel generators responding to a valid demand. This was done to determine if the performance of the emergency diesel generators (which are more than 25 years old) has deteriorated. The conditional probabilities of the diesel generators were computed and trended for the period from May 1990 to December 1992. The calculations indicate that the performance of the emergency diesel generators has not deteriorated in recent years, i.e., the conditional probabilities of the emergency diesel generators have been fairly stable over the last few years. This information will be one factor than may be considered in the decision to replace the emergency diesel generators

  10. FFTF [Fast Flux Test Facility] reactor shutdown system reliability reevaluation

    International Nuclear Information System (INIS)

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations

  11. Reactor cover gas monitoring at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bechtold, R.A.; Holt, F.E.; Meadows, G.E.; Schenter, R.E.

    1986-09-01

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the US Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100-day operating cycle began in April 1982 and the eighth operating cycle was completed in July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification

  12. Reactor water level measuring device

    International Nuclear Information System (INIS)

    Kuroki, Reiji; Asano, Tamotsu.

    1996-01-01

    A condensation vessel is connected to the upper portion of a reactor pressure vessel by way of a pipeline. The lower portion of the condensation vessel is connected to a low pressure side of a differential pressure transmission device by way of a reference leg pipeline. The high pressure side of the differential pressure transmission device is connected to the lower portion of the pressure vessel by way of a pipeline. The condensation vessel is equipped with a temperature sensor. When a temperature of a gas phase portion in the condensation vessel is lowered below a predetermined level, and incondensible gases in the condensation vessel starts to be dissolved in water, signals are sent from the temperature sensor to a control device and a control valve is opened. With such a constitution, CRD driving water flows into the condensation vessel, and water in which gases at the upper portion of the condensation vessel is dissolved flows into the pressure vessel by way of a pipeline. Then, gases dissolved in a reference water column in the reference leg pipeline are eliminated and the value of a reference water pressure does not change even upon abrupt lowering of pressure. (I.N.)

  13. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

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

  14. Dosimetry of mixed gamma - neutron fluxes in the active zone of working reactor and gamma-flux after quenching

    International Nuclear Information System (INIS)

    Mussaeva, M.A.; Zinov'ev, V.; Ibragimova, E.M.; Muminov, M.I.

    2006-01-01

    Full text: For carrying out experiments in the channels of nuclear reactor, it is necessary to know the distribution of neutron flux and the intensity of accompanying gamma-radiation both in the working and quenched regimes. Dosimetric parameter of transparent dielectrics is based on the effect of monotonous changing of optical absorption or luminescence under neutrons and/or gamma-radiation. While the radioactivity induced in an element monitor is proportional only to a neutron fluence beginning from a threshold energy. Therefore the aim of this work was to determine the values of neutron and gamma-component fluxes separately and evaluate the contribution of each into the defect production in dielectrics. We used very pure quartz glass of KU-1 type, produced in Russian State Optical Institute by fusion from SiCl 4 in the mixed flow of O 2 +H 2 (impurities of Cl and OH up to 10 -2 % and the rest - below 10 -4 %), SiO 2 glasses with 30 % Ba, and also pure Ni wire. Since under irradiation in the working reactor samples were undergone mixed neutron and gamma fluxes, we suggested determination of intensity of gamma-radiation from radio-nuclides (products of uranium fission) after quenching the reactor by the current of ionization chamber and glass dosimeters. Samples of SiO 2 -BaO together with Ni monitors were irradiated for 1 hour in 18 channels of the active zone of the working reactor both in the sealed ampoules and in the contact with water of the 1-st cooling circuit at 40 deg C. The linear dependence of the induced optical density on the absorbed dose of n 0 + γ-radiation was obtained. Ni -monitors not sensitive to γ-radiation gained the induced radioactivity proportional to the absorbed energy of neutron flux above 1 MeV. Neutron fluxes in the 18 channels varied from 9.53·10 11 to 1.21·10 13 cm -2 s -1 corresponding to fluences from 3.43·10 15 to 4.3·10 16 cm -2 . Optical density of band 215 nm ascertained to E ' - center, which is ≡ Si * near oxygen

  15. High Flux Isotope Reactor system RELAP5 input model

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  16. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Taryo, Taswanda

    2002-01-01

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

  17. Reactor cover gas monitoring at the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, R A; Holt, F E; Meadows, G E; Schenter, R E [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    The Fast Flux Test Facility (FFTF) is a 400 megawatt (thermal) sodium cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the U. S. Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100 day operating cycle began in April 1982 and the eighth operating cycle was completed In July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification. A liquid argon Dewar system provides the large volume of inert gas required for operation of the FFTF. The gas is used as received and is not recycled. Low concentrations of krypton and xenon in the argon supply are essential to preclude interference with the gas tag system. Gas chromatography has been valuable for detection of inadvertent air in leakage during refueling operations. A temporary system is installed over the reactor during outages to prevent oxide formation in the sodium vapor traps upstream from the on line gas chromatograph. On line gas monitoring by gamma spectrometry and grab sampling with GTSTs has been successful for the identification of numerous radioactive gas releases from creep capsule experiments as well as 9 fuel pin ruptures. A redundant fission gas monitoring system has been installed to insure constant surveillance of the reactor cover gas.

  18. Presentation of the High-Flux Reactor of the Institut Laue-Langevin

    International Nuclear Information System (INIS)

    Guyon, H.

    2006-01-01

    Full text of publication follows: The High-Flux Reactor (HFR) of the Institut Laue-Langevin is the world's most intense source of neutrons for fundamental research. Thanks to its extremely compact core, which is made up of a single fuel element, the HFR is capable of producing a neutron flux of up to 1.5.10 15 n.cm -2 .s -1 with a moderate power output of 58 MW. Its heavy water reflector thermalizes these neutrons, giving them a wave length of the order of one angstrom. They then become an excellent tool for exploring the atomic structure of matter. In order to provide a full neutron spectrum, the reactor is also equipped with a hot source (a block of graphite heated to 2000 deg. C) and two cold sources (a volume of liquid deuterium at 25 K). All the reactor's components can be replaced and adapted in order to keep pace with both changing scientific needs and changing safety requirements. For example, in 1992 the reactor block was replaced, a second cold source was installed in 1985, and the beam tubes are replaced at regularly intervals and are also occasionally modified. In the same way, the reactor's civil engineering structures are currently being reinforced in order to comply with the reassessment of the reference earthquake spectra. Finally, the Institut Laue-Langevin's reactor is equipped with three solid containment barriers: - the fuel cladding: during the 35 years the reactor has been in operation, a cladding failure has never been detected; - the leak-tight primary cooling system: this is partly submerged in a pool which provides radiological shielding; - the double-wall containment: an overpressure of air is maintained between the inner reinforced concrete wall and the outer metal wall. The High-Flux Reactor is therefore all set to provide the scientific community with top quality service for the next 20 years to come, on a site which: - is home to the brightest synchrotron in the world (ESRF); - benefits from the microbiology expertise of the EMBL

  19. Presentation of the High-Flux Reactor of the Institut Laue-Langevin

    Energy Technology Data Exchange (ETDEWEB)

    Guyon, H. [Institut Laue-Langevin, Grenoble (France)

    2006-07-01

    Full text of publication follows: The High-Flux Reactor (HFR) of the Institut Laue-Langevin is the world's most intense source of neutrons for fundamental research. Thanks to its extremely compact core, which is made up of a single fuel element, the HFR is capable of producing a neutron flux of up to 1.5.10{sup 15} n.cm{sup -2}.s{sup -1} with a moderate power output of 58 MW. Its heavy water reflector thermalizes these neutrons, giving them a wave length of the order of one angstrom. They then become an excellent tool for exploring the atomic structure of matter. In order to provide a full neutron spectrum, the reactor is also equipped with a hot source (a block of graphite heated to 2000 deg. C) and two cold sources (a volume of liquid deuterium at 25 K). All the reactor's components can be replaced and adapted in order to keep pace with both changing scientific needs and changing safety requirements. For example, in 1992 the reactor block was replaced, a second cold source was installed in 1985, and the beam tubes are replaced at regularly intervals and are also occasionally modified. In the same way, the reactor's civil engineering structures are currently being reinforced in order to comply with the reassessment of the reference earthquake spectra. Finally, the Institut Laue-Langevin's reactor is equipped with three solid containment barriers: - the fuel cladding: during the 35 years the reactor has been in operation, a cladding failure has never been detected; - the leak-tight primary cooling system: this is partly submerged in a pool which provides radiological shielding; - the double-wall containment: an overpressure of air is maintained between the inner reinforced concrete wall and the outer metal wall. The High-Flux Reactor is therefore all set to provide the scientific community with top quality service for the next 20 years to come, on a site which: - is home to the brightest synchrotron in the world (ESRF); - benefits from the

  20. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  1. Application of expert systems to heat exchanger control at the 100-megawatt high-flux isotope reactor

    International Nuclear Information System (INIS)

    Clapp, N.E. Jr.; Clark, F.H.; Mullens, J.A.; Otaduy, P.J.; Wehe, D.K.

    1985-01-01

    The High-Flux Isotope Reactor (HFIR) is a 100-MW pressurized water reactor at the Oak Ridge National Laboratory. It is used to produce isotopes and as a source of high neutron flux for research. Three heat exchangers are used to remove heat from the reactor to the cooling towers. A fourth heat exchanger is available as a spare in case one of the operating heat exchangers malfunctions. It is desirable to maintain the reactor at full power while replacing the failed heat exchanger with the spare. The existing procedures used by the operators form the initial knowledge base for design of an expert system to perform the switchover. To verify performance of the expert system, a dynamic simulation of the system was developed in the MACLISP programming language. 2 refs., 3 figs

  2. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  3. High performance light water reactor

    International Nuclear Information System (INIS)

    Squarer, D.; Schulenberg, T.; Struwe, D.; Oka, Y.; Bittermann, D.; Aksan, N.; Maraczy, C.; Kyrki-Rajamaeki, R.; Souyri, A.; Dumaz, P.

    2003-01-01

    The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

  4. Experimental studies on mitigation of LOCA for a high flux research reactor

    International Nuclear Information System (INIS)

    Saxena, A.K.

    2006-01-01

    Experimental studies on the rewetting behaviour of hot vertical annular channels were performed to study the mitigation of consequences of loss of coolant accident (LOCA) for a high flux research reactor. Studies were carried out to study the rewetting behaviour with hot inner tube, for bottom flooding and top flow rewetting conditions. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 litres per minute (l pm) (11.7 x 10 -5 m 3 s -1 ). The initial surface temperature of the inner tube was varied from 200 to 500 degC. (author)

  5. Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

    International Nuclear Information System (INIS)

    Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

    1987-01-01

    Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

  6. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  7. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

    1985-01-01

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  8. Critical Heat Flux Experiments on the Reactor Vessel Wall Using 2-D Slice Test Section

    International Nuclear Information System (INIS)

    Jeong, Yong Hoon; Chang, Soon Heung; Baek, Won-Pil

    2005-01-01

    The critical heat flux (CHF) on the reactor vessel outer wall was measured using the two-dimensional slice test section. The radius and the channel area of the test section were 2.5 m and 10 cm x 15 cm, respectively. The flow channel area and the heater width were smaller than those of the ULPU experiments, but the radius was greater than that of the ULPU. The CHF data under the inlet subcooling of 2 to 25 deg. C and the mass flux 0 to 300 kg/m 2 .s had been acquired. The measured CHF value was generally slightly lower than that of the ULPU. The difference possibly comes from the difference of the test section material and the thickness. However, the general trend of CHF according to the mass flux was similar with that of the ULPU. The experimental CHF data were compared with the predicted values by SULTAN correlation. The SULTAN correlation predicted well this study's data only for the mass flux higher than 200 kg/m 2 .s, and for the exit quality lower than 0.05. The local condition-based correlation was developed, and it showed good prediction capability for broad quality (-0.01 to 0.5) and mass flux ( 2 .s) conditions with a root-mean-square error of 2.4%. There were increases in the CHF with trisodium phosphate-added water

  9. Water level monitoring device in nuclear reactor

    International Nuclear Information System (INIS)

    Miura, Kiyohide; Otake, Tomohiro.

    1988-01-01

    Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

  10. Reactor water clean-up device

    International Nuclear Information System (INIS)

    Tanaka, Koji; Egashira, Yasuo; Shimada, Fumie; Igarashi, Noboru.

    1983-01-01

    Purpose: To save a low temperature reactor water clean-up system indispensable so far and significantly simplify the system by carrying out the reactor water clean-up solely in a high temperature reactor water clean-up system. Constitution: The reactor water clean-up device comprises a high temperature clean-up pump and a high temperature adsorption device for inorganic adsorbents. The high temperature adsorption device is filled with amphoteric ion adsorbing inorganic adsorbents, or amphoteric ion adsorbing inorganic adsorbents and anionic adsorbing inorganic adsorbents. The reactor water clean-up device introduces reactor water by the high temperature clean-up pump through a recycling system to the high temperature adsorption device for inorganic adsorbents. Since cations such as cobalt ions and anions such as chlorine ions in the reactor water are simultaneously removed in the device, a low temperature reactor water clean-up system which has been indispensable so far can be saved to realize the significant simplification for the entire system. (Seki, T.)

  11. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  12. High-flux first-wall design for a small reversed-field pinch reactor

    International Nuclear Information System (INIS)

    Cort, G.E.; Graham, A.L.; Christensen, K.E.

    1982-01-01

    To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high-technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time-averaged heat flux of 4.5 MW/m 2 with a conservatively estimated transient peak approximately twice the average value. We present the design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities. We combined a thermal analysis with a structural fatigue analysis to design the heat transfer module to last 10 6 cycles or one year at 80% duty for a 26-s power cycle. This fatigue life is compatible with a radiation damage life of 14 MW/yr/m 2

  13. Dosimetry issues for an ultra-high flux beam and multipurpose research reactor design

    International Nuclear Information System (INIS)

    West, C.D.

    1993-01-01

    The Advanced Neutron Source is a new user facility for all fields of neutron research, including neutron beam experiments, materials analysis, materials testing, and isotope production. The complement and layout of the experimental facilities have been determined sufficiently, at a conceptual design level, to make reliable cost and schedule estimates. The source of neutrons will be a heavy water reactor, constructed largely of aluminum, with an available thermal neutron flux 5--10 times higher than existing research reactors. Among the dosimetry issues to be faced are damage prediction and surveillance for component life attainment; measurement of fluence and spectra in regions where both change substantially over a distance of a few centimeters; and characterization and measurement of the radiation field in the research areas around the neutron beam experiments

  14. Measurement and calculation of fast neutron flux in a zero-energy reactor

    International Nuclear Information System (INIS)

    Day, D.H.; Fox, W.N.; Hyder, H.R.

    1963-05-01

    An activation technique for measuring relative fast neutron fluxes is described which has some advantages over the normal method using U238 fission. The technique is based on the formation of Rh 103 after inelastic scattering of neutrons above 100 keV in energy. This isomer decays with a 57.4 minute half-life giving an easily measurable γ-activity. The energy dependence of the inelastic scattering cross-section of Rh 103 is similar to that of the fission cross-section of U 238 thus making the results of direct relevance to reactor calculations. Using the Rh 103 activation technique, measurements have been made of the fast neutron flux distribution in a typical pressure tube heavy water lattice and are compared in this report with theoretical calculations using the MONTE CARLO method. (author)

  15. Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Bozic, M.; Zagar, T.; Ravnik, M.

    2001-01-01

    The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

  16. Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

    International Nuclear Information System (INIS)

    Scharmer, K.

    1969-01-01

    The results of experiments in the light water cooled D 2 O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k eff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D 2 O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [fr

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

    International Nuclear Information System (INIS)

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

    1998-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-05-01

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

  19. Entropy Generation Minimization for Reverse Water Gas Shift (RWGS Reactors

    Directory of Open Access Journals (Sweden)

    Lei Zhang

    2018-05-01

    Full Text Available Thermal design and optimization for reverse water gas shift (RWGS reactors is particularly important to fuel synthesis in naval or commercial scenarios. The RWGS reactor with irreversibilities of heat transfer, chemical reaction and viscous flow is studied based on finite time thermodynamics or entropy generation minimization theory in this paper. The total entropy generation rate (EGR in the RWGS reactor with different boundary conditions is minimized subject to specific feed compositions and chemical conversion using optimal control theory, and the optimal configurations obtained are compared with three reference reactors with linear, constant reservoir temperature and constant heat flux operations, which are commonly used in engineering. The results show that a drastic EGR reduction of up to 23% can be achieved by optimizing the reservoir temperature profile, the inlet temperature of feed gas and the reactor length simultaneously, compared to that of the reference reactor with the linear reservoir temperature. These optimization efforts are mainly achieved by reducing the irreversibility of heat transfer. Optimal paths have subsections of relatively constant thermal force, chemical force and local EGR. A conceptual optimal design of sandwich structure for the compact modular reactor is proposed, without elaborate control tools or excessive interstage equipment. The results can provide guidelines for designing industrial RWGS reactors in naval or commercial scenarios.

  20. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

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

  1. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

  2. Neutron scattering at the high-flux isotope reactor

    International Nuclear Information System (INIS)

    Cable, J.W. Chakoumakos, B.C.; Dai, P.

    1995-01-01

    The title facilities offer the brightest source of neutrons in the national user program. Neutron scattering experiments probe the structure and dynamics of materials in unique and complementary ways as compared to x-ray scattering methods and provide fundamental data on materials of interest to solid state physicists, chemists, biologists, polymer scientists, colloid scientists, mineralogists, and metallurgists. Instrumentation at the High- Flux Isotope Reactor includes triple-axis spectrometers for inelastic scattering experiments, a single-crystal four diffractometer for crystal structural studies, a high-resolution powder diffractometer for nuclear and magnetic structure studies, a wide-angle diffractometer for dynamic powder studies and measurements of diffuse scattering in crystals, a small-angle neutron scattering (SANS) instrument used primarily to study structure-function relationships in polymers and biological macromolecules, a neutron reflectometer for studies of surface and thin-film structures, and residual stress instrumentation for determining macro- and micro-stresses in structural metals and ceramics. Research highlights of these areas will illustrate the current state of neutron science to study the physical properties of materials

  3. Extraction of gadolinium from high flux isotope reactor control plates

    International Nuclear Information System (INIS)

    Kohring, M.W.

    1987-04-01

    Gadolinium-153 is an important radioisotope used in the diagnosis of various bone disorders. Recent medical and technical developments in the detection and cure of osteoporosis, a bone disease affecting an estimated 50 million people, have greatly increased the demand for this isotope. The Oak Ridge National Laboratory (ORNL) has produced 153 Gd since 1980 primarily through the irradiation of a natural europium-oxide powder followed by the chemical separation of the gadolinium fraction from the europium material. Due to the higher demand for 153 Gd, an alternative production method to supplement this process has been investigated. This process involves the extraction of gadolinium from the europium-bearing region of highly radioactive, spent control plates used at the High Flux Isotope Reactor (HFIR) with a subsequent re-irradiation of the extracted material for the production of the 153 Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (≥60% enriched in 152 Gd) resides in the europium-bearing region of the HFIR control components of which 70% is recoverable. At a specific activity yield of 40 curies of 153 Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed

  4. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  5. Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire. [Serbo-Croat] Ovaj izvestaj sadrzi sledece referate: Fluks termalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks epitermalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA; Fluks termalnih neurona u termalnoj koloni i bioloskom eksperimentalnom kanalu; Neutronska merenja u elementarnoj celiji reaktora RA; Temperaturni koeficijent reaktivnosti reaktora RA; Projekat uredjaja za merenje radioaktivnosti zice.

  6. Nuclear calculation methods for light water moderated reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1961-02-01

    This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)

  7. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  8. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  9. Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

    2009-01-01

    The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

  10. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  11. Method of operating heavy water moderated reactors

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1980-01-01

    Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

  12. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  13. Light water reactor safety research project

    International Nuclear Information System (INIS)

    Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

    1980-07-01

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  14. Boiling water reactor fuel bundle

    International Nuclear Information System (INIS)

    Weitzberg, A.

    1986-01-01

    A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods

  15. Consequences of corrosion of the high flux reactor channels

    International Nuclear Information System (INIS)

    1987-01-01

    The effects of corrosion can increase the probability of the channel losing its seal. In case of a slow leak, the phenomena happening can be considered as quasi-static. The closing of the safety valve takes place even before the leak water reaches the level of the exit window. In case of a fast leak in the case of helium filled channels, the dynamic effects are limited to the front part of the plug. As for the back part of the plug and the housing/safety valve unit, the consequences of a fast leak can be assimilated to those of a slow leak. This paper evaluates the results of an incident such as this for the reactor and the surrounding experimental zones

  16. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  17. Seismic strengthening of the ILL High Flux Reactor building

    International Nuclear Information System (INIS)

    Germane, Lionel; Plewinski, Francois; Thiry, Jean-Michel

    2006-01-01

    The Institut Max von Laue - Paul Langevin is an international research organisation and world leader in neutron science and technology. Since 1971 it has been operating the ILL HFR (High-Flux Reactor), the most intense continuous neutron source in the world. The ILL is governed by an international cooperation agreement between France, Germany and the United Kingdom; the fourth ten-year extension to the agreement was signed at the end of 2002, thus ensuring that the Institute will continue to operate until at least the end of 2013. In 2002 the facility underwent a general safety review, including an assessment of the impact of a safe shutdown earthquake. A broader programme for upgrading the installations and improving safety levels is now under way. As this has been treated in another paper, we will focus here on the seismic study carried out on the reactor building. The paper has the following contents: 1. Context; 1.1. Presentation of the ILL; 1.2. Description of the installations; 1.3. Safety objectives in the event of an earthquake; 1.4. Safety functions to be guaranteed in the event of an earthquake; 1.5. Safety functions required of the building; 2. Description of the building; 3. Organisation of the project; 3.1. Background; 3.2. Organisation; 4. General Methodology of the studies; 5. Progress of the studies; 5.1. Definition of the strengthening measures; 5.2. Validation of the strengthening option; 6. Seismic strengthening of the building; 6.1. Description of the strengthening measures; 6.2. Implementation of the strengthening measures; 6.2.1. Pilot operation; 6.2.2. Main operation; 7. Conclusion. To summarize, the presence of specialists in the ILL team, and the fact that the initial studies were performed by the project team itself, improved our general understanding of the issues and facilitated dialogue and exchange between all those involved (operators, technicians, outside experts, technical contractors and the French safety authorities). Everyone was

  18. On-line fast flux measurements in the BR2 reactor

    International Nuclear Information System (INIS)

    Vermeeren, L.

    2009-01-01

    Since 2001, CEA-Cadarache and the Belgian Nuclear Research Centre SCK-CEN are collaborating on the development and in-pile qualification of subminiature fission chambers (diameter of 1.5 mm). Initially, efforts concentrated on fission chambers for the in-pile measurement of thermal fluxes (with 235 U as fissile material). Meanwhile successful long-term tests of the prototypes have been performed in various environments: in low temperature (40-100 degress Celsius) BR2 pool water (up to a thermal neutron fluence of 3 1 0 21 n/cm 2 ) and in the CALLISTO PWR loop (300 degrees Celsius, 155 bars). The long-term qualification of derived industrial detectors (Photonis CFUZ53) in CALLISTO is still ongoing. However, for various types of irradiations in research reactors, the knowledge of the evolution of the fast neutron flux is even of more interest than the thermal flux data. Therefore the collaboration program was extended to the development and the in-pile qualification of subminiature or miniature fission chambers (with 3 mm diameter) for fast neutron detection, for which 242 Pu was selected as the optimal fissile material. In order to achieve the on-line in-pile measurement of fast neutron flux, the fission chambers will be operated in the Campbelling mode (based on the mean square fluctuation of the detector current). In this mode the gamma induced contribution to the signal can be efficiently suppressed. Moreover, a data processing software will take into account the evolution of the fissile deposit in order to assess on-line the fast flux sensitivity and to correct for the low energy neutron contributions. The final objective is to qualify a Fast Neutron Detector System (FNDS) able to provide on-line data for local fast neutron fluxes in Material Testing Reactors. The on-line measurement of the fast neutron flux would contribute significantly to the characterization of the irradiation conditions during test experiments with materials and innovative fuel elements

  19. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  20. Localized corrosion problems in water reactors

    International Nuclear Information System (INIS)

    Coriou, Henri.

    1977-01-01

    Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

  1. High flux isotope reactor cold source preconceptual design study report

    International Nuclear Information System (INIS)

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E.

    1995-12-01

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH 2 moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project

  2. Water feeding method upon reactor isolation

    International Nuclear Information System (INIS)

    Sasaki, Koichi; Takahara, Kuniaki; Hamamura, Kenji; Arakawa, Masahiro.

    1990-01-01

    The present invention concerns a method of feeding water upon reactor isolation in a plural loop type reactor having a plurality of reactor cooling systems. Water can be injected to a plurality of pools even if the pressure between the pools is not balanced and the water level in the reactor cooling system is optimally controlled. That is, water can be injected in accordance with the amount required for each of the pools by setting the opening of a turbine inlet steam control valve to somewhat higher than the cooling system pressure of the highest pressure loop. Water feeding devices upon reactor isolation were required by the same number as that for the reactor cooling systems. Whereas since pumps and turbines are used in common without worsening the water injection controllability to each of the loops according to the method of this invention and, accordingly, the cost performance can be improved. Further, since the opening degree of the turbine inlet steam control valve is controlled while making the difference pressure constant between the turbine inlet pressure and the pump exhaust pressure, the amount of the turbine exhausted steams can be reduced and, further, water injection controllability of the flow rate control valve in the injection line is improved. (I.S.)

  3. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  4. Reactor water quality degradation suppressing method upon reactor start up

    International Nuclear Information System (INIS)

    Maeda, Katsuharu.

    1993-01-01

    Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

  5. Water chemistry in WWER reactors

    International Nuclear Information System (INIS)

    Yurmanov, V.A.; Mamet, V.A.; Shestakov, Yu.M.; Amosov, M.M.

    1997-01-01

    In this paper ''Water Chemistry in WWER Reactors'', are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH T values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ''soft decontamination'' involving changing the KOH concentration and, hence, the pH T before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs

  6. Water chemistry in WWER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yurmanov, V A; Mamet, V A; Shestakov, Yu M; Amosov, M M [All-Russian Scientific Research Inst. for Nuclear Power Plants Operation, Moscow (Russian Federation)

    1997-02-01

    In this paper ``Water Chemistry in WWER Reactors``, are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH{sub T} values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ``soft decontamination`` involving changing the KOH concentration and, hence, the pH{sub T} before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs.

  7. Physics of pressurized water reactors

    International Nuclear Information System (INIS)

    Gruen, A.

    1980-01-01

    The objective of this lecture is to demonstrate typical problems and solutions encountered in the design and operation of PWR power plants. The examples selected for illustration refer to PWR's of KWU design and to results of KWU design methods. In order to understand the physics of a power reactor it is necessary to have some knowledge of the structure and design of the power plant system of which the reactor is a part. It is therefore assumed that the reader is familiar with the design of the more important components and systems of a PWR, such as fuel assemblies, control assemblies, core lay-out, reactor coolant system, instrumentation. (author)

  8. Water-immersion type ship reactor

    International Nuclear Information System (INIS)

    Okada, Hiroki; Yamamura, Toshio.

    1996-01-01

    In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

  9. Emergency cooling of presurized water reactor

    International Nuclear Information System (INIS)

    Sykora, D.

    1981-01-01

    The method described of emergency core cooling in the pressurized water reactor is characterized by the fact that water is transported to the disturbed primary circuit or direct to the reactor by the action of the energy and mass of the steam and/or liquid phase of the secondary circuit coolant, which during emergency core cooling becomes an emergency cooling medium. (B.S.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-12-01

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

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

    International Nuclear Information System (INIS)

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

    2004-12-01

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

  12. Transmutation of technetium into stable ruthenium in high flux conceptual research reactor

    International Nuclear Information System (INIS)

    Amrani, N.; Boucenna, A.

    2007-01-01

    The effectiveness of transmutation for the long lived fission product technetium-99 in high flux research reactor, considering its large capture cross section in thermal and epithermal region is evaluated. The calculation of Ruthenium concentration evolution under irradiation was performed using Chain Solver 2.20 code. The approximation used for the transmutation calculation is the assumption that the influence of change in irradiated materials structures on the reactor operator mode characteristics is insignificant. The results on Technetium transmutation in high flux research reactor suggested an effective use of this kind of research reactors. The evaluation brings a new concept of multi-recycle Technetium transmutation using HFR T RAN (High Flux Research Reactor for Transmutation)

  13. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

    EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

  14. Radiological characterization for small type light water reactor

    International Nuclear Information System (INIS)

    Tanaka, Ken-ichi; Ichige, Hideaki; Tanabe, Hidenori

    2011-01-01

    In order to plan a decommissioning, amount investigation of waste materials and residual radioactivity inventory evaluation must be performed at the first stage of preparatory tasks. These tasks are called radiological characterization. Reliable information from radiological characterization is crucial for specification of decommissioning plan. With the information, we can perform radiological safety analysis and optimize decommissioning scenario. Japan Atomic Power Company (JAPC) has already started preparatory tasks for Tsuruga Nuclear Power Plant Unit 1 (TS-1) that is the first commercial Small Type Light Water Reactor in Japan. To obtain reliable information about residual radioactivity inventory, we improved radioactivity inventory evaluation procedure. The procedure consists of neutron flux distribution calculation and radioactivity distribution calculation. We need a better understanding about characteristics of neutron transport phenomena in order to obtain reliable neutron flux distribution. Neutron flux was measured in Primary Containment Vessel (PCV) at 30 locations using activation foils. We chose locations where characteristic phenomena can be observed. Three dimensional (3D) neutron flux calculation was also performed to simulate continuous changes of neutron flux distribution. By assessing both the measured values and 3D calculation results, we could perform the calculation that simulates the phenomena well. We got knowledge about how to perform an appropriate neutron flux distribution calculation and also became able to calculate a reliable neutron flux distribution. Using the neutron flux distribution, we can estimate a reliable radioactivity distribution. We applied network-parallel-computing method to the estimation. And further we developed 'flux level approximation method' which use linear or parabola fitting method to estimation. Using these new methods, radioactivity by neutron irradiation, which is radioisotope formation, was calculated at

  15. Environmental effects on the response of self-powered flux detectors in CANDU reactors

    International Nuclear Information System (INIS)

    Lynch, G.F.; Shields, R.B.; Joslin, C.W.

    1976-01-01

    Self-powered flux detectors are playing an increasingly important role in the control and safety systems of CANDU-type reactors. In this paper we report on recent experiments to determine how local reactor conditions affect the output signals from self-powered detectors with vanadium, platinum and cobalt emitters. The results are interpreted in terms of variations in the local neutron, γ-ray and electron fluxes. (author)

  16. Studies on the instrumentation of a beam-tube medium flux reactor

    International Nuclear Information System (INIS)

    Axmann, A.; Pollet, J.L.; Queudot, J.

    1979-01-01

    In the years 1977/78, the ad hoc commitee for medium-flux reactor development of the Federal Ministry for Research and Technology developed constructional concepts for a medium-flux reactor to be utilized by beam tube experiments. The HMI has elaborated contributions for discussions of the subject of instrumentation, in particular for experiments in solid state physics. These contributions are contained in the report. (orig./RW) [de

  17. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    International Nuclear Information System (INIS)

    1989-07-01

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

  18. Pressurised water reactor in the UK

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    Since the Three Mile Island accident there has been much debate about the safety considerations of Pressurised Water Reactors. Their development will continue throughout the world but it will be based upon the lessons learned from that unfortunate accident. In the United Kingdom there is a public enquiry discussing all aspects of the reactor. The papers given in this book provide an informed addition to the literature. The design, safety and licensing and construction of a pressurised water reactor system are discussed in detail. Considerations stemming from the Three Mile Island accident are presented

  19. Hydriding failure in water reactor fuel elements

    International Nuclear Information System (INIS)

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  20. Water injection device for reactor container

    International Nuclear Information System (INIS)

    Sakaki, Isao.

    1996-01-01

    A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

  1. Emergency water supply facility for nuclear reactor

    International Nuclear Information System (INIS)

    Karasawa, Toru

    1998-01-01

    Water is stored previously in an equipment storage pit disposed on an operator floor of a reactor building instead of a condensate storage vessel. Upon occurrence of an emergency, water is supplied from the equipment storage pit by way of a sucking pipeline to a pump of a high pressure reactor core water injection circuit and a pump of a reactor-isolation cooling circuit to supply water to a reactor. The equipment storage pit is arranged in a building so that the depth thereof is determined to keep the required amount of water by storing water at a level lower than the lower end of a pool gate during normal operation. Water is also supplied from the equipment storage pit by way of a supply pipeline to a spent fuel storage pool on the operation floor of the reactor building. Namely, water is supplied to the spent fuel storage pool by a pump of a fuel pool cooling and cleaning circuit. This can eliminate a suppression pool cleaning circuit. (I.N.)

  2. Power spectral density measurements with 252Cf for a light water moderated research reactor

    International Nuclear Information System (INIS)

    King, W.T.; Mihalczo, J.T.

    1979-01-01

    A method of determining the reactivity of far subcritical systems from neutron noise power spectral density measurements with 252 Cf has previously been tested in fast reactor critical assemblies: a mockup of the Fast Flux Test Facility reactor and a uranium metal sphere. Calculations indicated that this measurement was feasible for a pressurized water reactor (PWR). In order to evaluate the ability to perform these measurements with moderated reactors which have long prompt neutron lifetimes, measurements were performed with a small plate-type research reactor whose neutron lifetime (57 microseconds) was about a factor of three longer than that of a PWR and approx. 50% longer than that of a boiling water reactor. The results of the first measurements of power spectral densities with 252 Cf for a water moderated reactor are presented

  3. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

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

  4. Water desalination using different capacity reactors options

    International Nuclear Information System (INIS)

    Alonso, G.; Vargas, S.; Del Valle, E.; Ramirez, R.

    2010-01-01

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity, cogeneration of potable water production and nuclear electricity is an option to be assessed. In this paper we will perform an economical comparison for cogeneration using a big reactor, the AP1000, and a medium size reactor, the IRIS, both of them are PWR type reactors and will be coupled to the desalination plant using the same method. For this cogeneration case we will assess the best reactor option that can cover both needs using the maximum potable water production for two different desalination methods: Multistage Flash Distillation and Multi-effect Distillation. (authors)

  5. New lineup of light water reactors

    International Nuclear Information System (INIS)

    Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

    2007-01-01

    Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

  6. Gravity Scaling of a Power Reactor Water Shield

    International Nuclear Information System (INIS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa n . These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined

  7. Spatial fluxes and energy distributions of reactor fast neutrons in two types of heat resistant concretes

    International Nuclear Information System (INIS)

    Akki, T.S.; Benayad, S.A.; Megahid, R.M.

    1992-01-01

    Measurements have been carried out to study the spatial fluxes and energy distributions of reactor fast neutrons transmitted through two types of heat resistant concretes, serpentine concrete and magnetic lemonite concrete. The physical, chemical and mechanical properties of these concretes were checked by well known techniques. In addition, the effect of heating at temperatures up to 500deg C on the crystaline water content was checked by the method of differential thermal analysis. Measurements were performed using a collimated beam of reactor neutrons emitted from a 10 MW research reactor. The neutron spectra transmitted through concrete barriers of different thickness were measured by a scintillation spectrometer with NE-213 liquid organic scintillator. Discrimination against undesired pulses due to gamma-rays was achieved by a method based on pulse shape discrimination technique. The operating principle of this technique is based on the comparison of two weighted time integrals of the detector signal. The measured pulse amplitude distribution was converted to neutron energy distribution by a computational code based on double differentiation technique. The spectrometer workability and the accuracy of the unfolding technique were checked by measuring the neutron spectra of neutrons from Pu-α-Be and 252 Cf neutron sources. The obtained neutron spectra for the two concretes were used to derive the total cross sections for neutrons of different energies. (orig.)

  8. Operating manual for the High Flux Isotope Reactor. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    None

    1965-06-01

    This report contains a comprehensive description of the High Flux Isotope Reactor facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procedures are presented in another report.

  9. Transport calculation of neutron flux distribution in reflector of PW reactor

    International Nuclear Information System (INIS)

    Remec, I.

    1982-01-01

    Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

  10. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    International Nuclear Information System (INIS)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

  11. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  12. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  13. Flux distribution by neutrons semi-conductors detectors during the startup of the EL4 reactor

    International Nuclear Information System (INIS)

    Fuster, S.; Tarabella, A.

    1967-01-01

    The Cea developed neutron semi-conductors detectors which allows a quasi-instantaneous monitoring of neutrons flux distribution, when placed in a reactor during the tests. These detectors have been experimented in the EL4 reactor. The experiment and the results are presented and compared with reference mappings. (A.L.B.)

  14. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated

  15. One procedure for determination of the neutron flux in the nuclear reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Bulovic, V; Krtil, J; Maksimovic, Z; Martinc, R [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1979-09-15

    Possibility of determination of the neutron flux in the fuel of a heavy water reactor has been examined. In determination of the flux an iterative procedure was used to compare calculated and measured contents of several fission products. The former contents were determined by calculation of the burning process balance and the latter by non-destructive gamma-spectrometric analysis of fuel. The obtained results prove the possibility of such determination of not only the average value of the flux but also of the change of its intensity during utilization of fuel (author) U radu je ispitivana mogucnost odredjivanja fluksa neutrona u gorivu teskovodnog nuklearnog reaktora. Pri odredjivanju fluksa koriscen je iterativni postupak, u kome se porede izracunati i izmereni sadrzaji nekoliko gama-radioaktivnih fisionih produkata. Prvi sadrzaji su odredjivani preko proracuna bilansa procesa sagorevanja, a drugi - nedestruktivnom gama-spektrometrijskom analizom goriva. Dobijeni rezultati potvrdjuju mogucnost ovakvog odredjivanja ne samo srednje vrednosti fluksa, vec i promene njegovog intenziteta u toku koriscenja goriva (author)

  16. Critical heat flux predictions for the Sandia Annular Core Research Reactor

    International Nuclear Information System (INIS)

    Rao, D.V.; El-Genk, M.S.

    1994-08-01

    This study provides best estimate predictions of the Critical Heat Flux (CHF) and the Critical Heat Flux Ratio (CHFR) to support the proposed upgrade of the Annual Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) from its present value of 2 MWt to 4 MWt. These predictions are based on the University of New Mexico (UNM) - CHF correlation, originally developed for uniformly heated vertical annuli. The UNM-CHF correlation is applicable to low-flow and low-pressure conditions, which are typical of those in the ACRR. The three hypotheses that examined the effect of the nonuniform axial heat flux distribution in the ACRR core are (1) the local conditions hypotheses, (2) the total power hypothesis, and (3) the global conditions hypothesis. These hypotheses, in conjunction with the UNM-CHF correlation, are used to estimate the CHF and CHFR in the ACRR. Because the total power hypothesis predictions of power per rod at CHF are approximately 15%-20% lower than those corresponding to saturation exit conditions, it can be concluded that the total power hypothesis considerably underestimates the CHF for nonuniformly heated geometries. This conclusion is in agreement with previous experimental results. The global conditions hypothesis, which is more conservative and more accurate of the other two, provides the most reliable predictions of CHF/CHFR for the ACRR. The global conditions hypothesis predictions of CHFR varied between 2.1 and 3.9, with the higher value corresponding to the lower water inlet temperature of 20 degrees C

  17. On Line Neutron Flux Mapping in Fuel Coolant Channels of a Research Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Villard, Jean-Francois; Destouches, Christophe; Braoudakis, George; Wassink, David; Sinclair, Bradley; Osborn, John-C.; Wu, Huayou; Blandin, C.; Thevenin, Mathieu; Corre, Gwenole; Normand, Stephane

    2013-06-01

    This work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E ∼0.97). (authors)

  18. Three-dimensional calculations of neutron streaming in the beam tubes of the ORNL HFIR [High Flux Isotope Reactor] Reactor

    International Nuclear Information System (INIS)

    Childs, R.L.; Rhoades, W.A.; Williams, L.R.

    1988-01-01

    The streaming of neutrons through the beam tubes in High Flux Isotope Reactor at Oak Ridge National Laboratory has resulted in a reduction of the fracture toughness of the reactor vessel. As a result, an evaluation of vessel integrity was undertaken in order to determine if the reactor can be operated again. As a part of this evaluation, three-dimensional neutron transport calculations were performed to obtain fluxes at points of interest in the wall of the vessel. By comparing the calculated and measured activation of dosimetry specimens from the vessel surveillance program, it was determined that the calculated flux shape was satisfactory to transpose the surveillance data to the locations in the vessel. A bias factor was applied to correct for the average C/E ratio of 0.69. 8 refs., 7 figs., 3 tabs

  19. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1987-01-01

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  20. Dissolution flowsheet for high flux isotope reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-01

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

  2. The effect of an iron plug on the neutron flux distributions in water

    International Nuclear Information System (INIS)

    Lotfi, A.; Maayouf, R.M.A.; Megahid, R.

    1978-01-01

    This work is concerned with studying both fast and thermal neutron fluxes distribution in water and its perturbation due to the presence of a cylindrical iron plug. The measurements were carried out using a collimated neutron beam emitted from one of the horizontal channels of the ET-RR-1 reactor. The fast neutron fluxes were measured using phosphorus activation detectors, while the thermal neutron ones were measured using fission fragment track detectors from glass. The results show that the presence of an iron plug causes a remarkable change in the intensities of both the fast and thermal neutron fluxes distribution in the water medium surrounding the iron plug. The flux intensities at the peaks, formed beyond the iron plug in case of thermal neutrons, are also compared with values calculated using the available emperical formula

  3. Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

    OpenAIRE

    Verma, Vasudha

    2017-01-01

    Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

  4. A conceptual high flux reactor design with scope for use in ADS ...

    Indian Academy of Sciences (India)

    By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to ...

  5. The effect of temperature and the control rod position on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Omar, H.; Ghazi, N.

    2007-01-01

    The effect of water and fuel temperature increase and changes in the control rod positions on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor (MNSR) is discussed. The cross sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the special neutron flux distribution using four energy groups. This work shows that water and fuel temperature increase in the reactor during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. Changing the control rod position does not affect as well the spatial neutron flux distribution except in the region around the control rod position. This stability in the spatial neutron flux distribution, especially in the inner and outer irradiation sites, makes MNSR as a good tool for the neutron activation analysis (NAA) technique and production of radioisotopes with medium or short half lives during the reactor daily operating time. (author)

  6. Mathematical modeling of water radiolysis in the Syrian MNSR reactor

    International Nuclear Information System (INIS)

    Soukieh, M.

    2009-11-01

    Because it is difficult to measure the concentration of the radiolytic species in reactors under operating conduction, they must be estimated by computer simulation techniques. This study discusses the mathematical modeling of water radiolysis modeling of the MNSR nuclear reactor cooling water. The mathematical model comprising of 13 differential equations describe 55 chemical reactions of radiolytic species e - a q H + , OH - , H, H 2 , OH, HO 2 , O 2 , HO - 2 , O - , O - 2 , O - 3 . The mathematical model have been tested and it shows a good agreement of the computed values in this work with the results cited in references [1,18] in case of only γray irradiation of pure water with dose rate of 1.18x10 19 eV/L s. The neutron fluxes and dose rates at the interface of cladding-water for the different fuel rings in the MNSR core are determined using MCNP-4C code. In addition, the time dependent of the radiolytic specie concentrations were estimated for max. and min. dose rates and at temperature of 20 degree centigrade in the MNSR. The radiolytic specie concentrations reach the steady sate after about 200-400 s. The radiolytic specie concentrations order of H 2 , O 2 , H 2 O 2 were about ppb. Also this study shows the possibility of suppressed the water radiolysis reactions by adding hydrogen to the MNSR reactor cooling water. (author)

  7. Method of measuring reactor water level

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1979-01-01

    Purpose: To provide a water level measuring system so that a reactor water level detecting signal can be corrected in correspondence to a recirculation flow, thereby to carry out a correct water level detection in a wide range of the reactor. Method: According to the operation record of a precursor reactor, the ratio Δh of the lowering of the water level due to the recirculation flow is lowered in proportion to the ratiowith respect to the rated differential pressure of the recirculation flow. Accordingly, the flow of recirculation pump is measured by an elbow differential pressure generator utilizing an elbow of a pipe, and the measured value is multiplied by a gain by a ratio setter, and therefter, an addition computation is carried out by an adder for correcting the signal from a water level detector. When the signal from the water level detector is corrected in this manner, the influence of the lowering of the water level due to the recirculation flow can be removed, and an interlocker predetermined in the defined water level can be actuated, thus the influence of the dynamic pressure due to the recirculation flow acting on the instrumental pipe line detecting the reactor water level can be removed effectively. (Yoshino, Y.)

  8. Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

    Energy Technology Data Exchange (ETDEWEB)

    Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

    2014-01-15

    Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

  9. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  10. Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Manas Diaz, L; Montes Ponce de leon, J.

    1960-07-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

  11. Requirements for light water reactors

    International Nuclear Information System (INIS)

    Hedin, F.

    2009-01-01

    The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10 -6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

  12. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  13. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    1988-10-01

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  14. Reliability of reactor plant water cleanup pumps

    International Nuclear Information System (INIS)

    Pearson, J.L.

    1979-01-01

    Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

  15. Review of issues relating to enhancement of critical heat flux in nuclear reactors - 15220

    International Nuclear Information System (INIS)

    Hasan, S.M.K.; Chowdhury, M.A.Z.; Sarkar, M.A.R.

    2015-01-01

    Critical heat flux (CHF) is manifested by sharp reduction in heat transfer coefficient and a drastic rise in fuel rod surface temperature, which may lead to failure of the fuel cladding and is of immense importance in reactor design and operation. In this paper, the existing literature on CHF is studied and analyzed. Different modeling techniques such as geometry modeling and fluid modeling are highlighted. CHF enhancement techniques such as swirl flow, porous coating, nano-fluids or mechanical vibrations are analyzed and their suitability for different applications is discussed. The advances in prediction methods for tubes and rod bundles are also presented. CHF correlations and their applications have been compiled to estimate the CHF for fluids other than water or other geometries. Finally, future research needs are identified and presented. (authors)

  16. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

    Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

    1964-01-01

    France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

  17. Utility requirements for advanced light water reactors

    International Nuclear Information System (INIS)

    Machiels, A.; Gray, S.; Mulford, T.; Rodwell, E.

    1996-01-01

    The nuclear energy industry is actively engaged in developing advanced light water reactor (ALWR) designs for the next century. The new designs take advantage of the thousands of reactor-years of experience that have been accumulated by operating over 400 plants worldwide. The EPRI effort began in the early 1980's, when a survey of utility executives was conducted to determine their prerequisites for ordering nuclear power plants. The results were clear: new plants had to be simpler and safer, and have greater design margins, i.e., be more forgiving. The utility executives also supported making improvements to the established light water reactor technology, rather than trying to develop new reactor concepts. Finally, they wanted the option to build mid-size plants (∼600 MWe) in addition to full-size plants of more than 1200 MWe. 4 refs

  18. Safety aspects of pressurised water reactors

    International Nuclear Information System (INIS)

    1985-01-01

    This submission to the Health and Safety Executive has been prepared by the Institution of Professional Civil Servants (IPCS) as a contribution to the debate on safety aspects associated with Pressurized Water Reactors (PWRs). Although supporting an energy policy which includes the development of nuclear power, assurances are sought on a number of safety issues if it is decided that this should be generated by a PWR-type reactor. These issues are listed. In particular the following are mentioned: the wider publication of design information, the use of elastic-plastic fracture mechanics as the basis for determining pressure vessel integrity, the failure rate of steam generating units, water coolant quality control, greater investigation of two-phase flow accident conditions, the components of the reactor cooling system and training of reactor personnel in the understanding of LOCA effects. (U.K.)

  19. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Sriram, Jayasree; Vijayan, K.; Kain, Vivekanad; Velmurugan, S.

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

  20. Nonlinear dynamics of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1983-01-01

    Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

  1. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  2. Steam explosions in light water reactors

    International Nuclear Information System (INIS)

    1981-01-01

    The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

  3. Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Israr, M.; Shami, Qamar-ud-din; Pervez, S.

    1997-11-01

    In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

  4. Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods

    International Nuclear Information System (INIS)

    Rothrock, R.B.

    1991-01-01

    Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs

  5. High Flux Isotope Reactor quarterly report, July--September 1975

    International Nuclear Information System (INIS)

    McCord, R.V.; Corbett, B.L.

    1975-01-01

    The replacement of the permanent beryllium reflector was completed this quarter. The reactor was shut down for 87 days for this maintenance operation. Erosion of the sealing surface at the stainless steel adaptor flange on the HB-1 beam tube facility was confirmed. A soft metallic O-ring was used to effect a seal when this facility was reassembled. A comprehensive inspection of the normally inaccessible parts of the reactor pressure vessel was made. No abnormalities were found

  6. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Nanis, R.

    2000-01-01

    Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

  7. Reactor water clean-up device

    International Nuclear Information System (INIS)

    Sawa, Toshio; Takahashi, Sankichi; Takashima, Yoshie.

    1983-01-01

    Purpose: To efficiently eliminate radioactive materials such as iron oxide and cobalt ions with less heat loss by the use of an electrode assembly applied with a direct current. Constitution: In a reactor water clean-up device adapted to pass reactor water through an electrode assembly comprising at least a pair of anode and cathode applied with a direct current to eliminate various types of ions contained in the reactor water by way of the electrolysis or charge neutralization at the anode, the cathode is constituted with a corrosion resistant grid-like or porous metal plate and a layer to the upper portion of the metal plate filled with a plurality of metal spheres of about 1 - 5 mm diameter, and the anode is made of insoluble porous or spirally wound metal material. (Seki, T.)

  8. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-09-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  9. Dissolution Flowsheet for High Flux Isotope Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas, allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  10. Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

  11. Analysis of neutron flux increase in the horizontal experimental channels of Ra reactor - masters thesis

    International Nuclear Information System (INIS)

    Strugar, P.

    1964-12-01

    Calculation and experimental results shown in this paper show that higher thermal neutron flux is obtained in the reactor core with central horizontal reflector at the same power level. The flux is increased when the moderation capability of the core is decreased. Apart from increase of the thermal component of the neutron flux in the experimental channels, the central reflector causes decrease of the epithermal neutron flux and gamma radiation intensity. This is very useful for studying (n, γ) reaction, neutron diffraction, etc. [sr

  12. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  13. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements

    International Nuclear Information System (INIS)

    Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-01-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. - Highlights: • Neutron flux redistribution due to control rod movement in JSI TRIGA has been studied. • Detector response sensitivity to the control rod position has been minimized. • Optimal radial and axial detector positions have been determined

  14. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

    1986-01-01

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

  15. ULTRA SCWR+: Practical advanced water reactor concepts

    International Nuclear Information System (INIS)

    Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

    2008-01-01

    Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

  16. Beyond the light water reactor

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1980-01-01

    One of the strong interests in examining alternative nuclear fuel cycles is to search for schemes that are more efficient than LWRs in their use of uranium, but that do not carry the additional proliferation hazard associated with widespread plutonium utilization. One possibility is to improve the uranium efficiency of current reactor types by other means than recycling. A second possibility, offering greater potential for improvement, is to utilize thorium-uranium fuel cycles in which uranium-233 is denatured by the addition of uranium-238, making enrichment necessary to yield weapons-usable material. The bulk of the reactor's fuel material would be thorium-232, so that most of the fissile material produced would be uranium-233. It is important to recognize that these two possibilities - once through improvements and denatured thorium-uranium - could be introduced sequentially in reactor types that are currently in use. Fuel cycles may change over time, but it is equally significant from the point of view of non-proliferation that they will also vary from place to place and, most importantly, from country to country. The author argues that alternative nuclear power systems and a slower growth may affect the diversion of nuclear materials to weapons. A real question, though, is whether we have time to explore the possibilities. It has become apparent that predictions made of the growth rate for nuclear power were too high. The 1000 large power plants the US was to have by the year 2000 have been reduced to fewer than 300. This reduces the pressure, resulting from our limited uranium resources, to push the LMFBR. Extra time gives us a chance to examine the possibilities

  17. Detection and location of can rupture in reactors cooled by a flow of water

    International Nuclear Information System (INIS)

    Le Meur, R.

    1968-01-01

    This report brings together the principal methods of fission-product detection used for water reactors. The position, type and method of adjustment is given for each detector. The methods for localizing the defective elements are explained, in particular those using water sampling or decreases in the flux. A few installations are briefly described. They correspond to particular types of reactors using boiling, pressurized or cold water. Amongst the many methods used, it can be noted that when the fuel is resistant, the installations are fairly compact. In nuclear super-heated reactors on the other hand, the study of fuel behaviour calls for larger installations. An identification of defective elements exists when the reactor structure allows it. If this is not possible, a localization in a group of elements is obtained by a flux depression. (author) [fr

  18. EPRI program in water reactor safety

    International Nuclear Information System (INIS)

    Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

    1975-01-01

    The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

  19. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

    This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

  20. The water chemistry of CANDU PHW reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-01-01

    This review will discuss the chemistry of the three major water circuits in a CANDU-PHW reactor, viz., the Primary Heat Transport (PHT) water, the moderator and the boiler water. An important consideration for the PHT chemistry is the control of corrosion and of the transport of corrosion products to minimize the growth of radiation fields. In new reactors the PHT will be allowed to boil, requiring reconsideration of the methods used to radiolytic oxygen and elevate the pH. Separation of the moderator from the PHT in the pressure-tubed CANDU design permits better optimization of the chemistry of each system, avoiding the compromises necessary when the same water serves both functions. Major objectives in moderator chemistry are to control (a) the radiolytic decomposition of D 2 0; (b) the concentration of soluble neutron poisons added to adjust reactivity; and (c) the chemistry of shutdown systems. The boiler water and its feed water are treated to avoid boiler tube corrosion, both during normal operation and when perturbations are caused to the feed by, for example, leaks in the condenser tubes which permit ingress of untreated condenser cooling water. Development of a system for automatic analysis and control of feed water to give rapid, reliable response to abnormal conditions is a novel feature which has been developed for incorporation in future CANDU-PHW reactors. (author)

  1. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  2. Safety aspects of water chemistry in light water reactors

    International Nuclear Information System (INIS)

    1988-12-01

    The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

  3. Anti-neutrino flux in a research reactor for non-proliferation application

    Energy Technology Data Exchange (ETDEWEB)

    Khakshournia, Samad; Foroughi, Shokoufeh [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Atomic Energy Organization of Iran (AEOI)

    2017-11-15

    Owing to growing interest in the study of emitted antineutrinos from nuclear reactors to test the Atomic Energy Agency safeguards, antineutrino flux was studied in the Tehran Research Reactor (TRR) using ORIGEN code. According to our prediction, antineutrino rate was obtained 2.6 x 10{sup 17} (v{sub e}/sec) in the core No. 57F of the TRR. Calculations indicated that evolution of antineutrino flux was very slow with time and the performed refueling had not an observable effect on antineutrino flux curve for a 5 MW reactor with the conventional refueling program. It is seen that for non-proliferation applications the measurement of the contribution of {sup 239}Pu to the fission using an antineutrino detector is not viable in the TRR.

  4. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Soppet, W.K.; Shack, W.J.

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  5. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  6. Assembly homogenization techniques for light water reactor analysis

    International Nuclear Information System (INIS)

    Smith, K.S.

    1986-01-01

    Recent progress in development and application of advanced assembly homogenization methods for light water reactor analysis is reviewed. Practical difficulties arising from conventional flux-weighting approximations are discussed and numerical examples given. The mathematical foundations for homogenization methods are outlined. Two methods, Equivalence Theory and Generalized Equivalence Theory which are theoretically capable of eliminating homogenization error are reviewed. Practical means of obtaining approximate homogenized parameters are presented and numerical examples are used to contrast the two methods. Applications of these techniques to PWR baffle/reflector homogenization and BWR bundle homogenization are discussed. Nodal solutions to realistic reactor problems are compared to fine-mesh PDQ calculations, and the accuracy of the advanced homogenization methods is established. Remaining problem areas are investigated, and directions for future research are suggested. (author)

  7. Heat diffusion in cylindrical fuel elements of water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1961-09-15

    This report contains a theoretical study of heat diffusion in the cylindrical fuel elements of water reactors. After setting up appropriate boundary conditions on the temperature, the steady state Fourier equation is solved both for a flat and a tilted fission power source. It is shown that source tilting does not have an appreciable effect on the peak fuel temperature while the heat flux to the coolant suffers a circumferential variation of less than a half of that of the fission power. In the last section, the theory is extended to include the effect of a flat, time dependent fission power. The time dependent Fourier equation is solved by means of a Dini series of Bessel functions which is shown to be rapidly convergent. From this series is derived expressions for the fuel element transfer functions required in reactor servo-analysis. These have the form of a rapidly convergent series of time-lag terms. (author)

  8. Historical perspective of thermal reactor safety in light water reactors

    International Nuclear Information System (INIS)

    Levy, S.

    1986-01-01

    A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective

  9. INDIAN POINT REACTOR REACTIVITY AND FLUX DISTRIBUTION MEASUREMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Batch, M. L.; Fischer, F. E.

    1963-11-15

    The reactivity of the Indian Point core was measured near zero reactivity at various shim and control rod patterns. Flux distribution measurements were also made, and the results are expressed in terms of power peaking factors and normalized detector response during rod withdrawal. (D.L.C.)

  10. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

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

  11. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  12. Determination of neutron flux densities in WWR-S reactor core

    International Nuclear Information System (INIS)

    Tomasek, F.

    1989-04-01

    The method is described of determining neutron flux densities and neutron fluences using activation detectors. The basic definitions and relations for determining reaction rates, fluence and neutron flux as well as the characteristics of some reactions and of sitable activation detectors are reported. The flux densities were determined of thermal and fast neutrons and of gamma quanta in the WWR-S reactor core. The data measured in the period 1984-1987 are tabulated. Cross sections for the individual reactions were determined from spectra measurements processed using program SAND-II and cross section library ENDF-B IV. Neutron flux densities were also measured for the WWR-S reactor vertical channels. (E.J.). 10 figs., 8 tabs., 111 refs

  13. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-12-01

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  14. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    Moore, Mark; George, Robert; Spence, Harry; Nguyen, John

    1992-01-01

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  15. Quality assurance plan, Westinghouse Water Reactor Divisions

    Energy Technology Data Exchange (ETDEWEB)

    1976-03-01

    The Quality Assurance Program used by Westinghouse Nuclear Energy Systems Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements.

  16. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  17. Facilitation of decommissioning light water reactors

    International Nuclear Information System (INIS)

    Moore, E.B. Jr.

    1979-12-01

    Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented. A wide range of facilitation methods - from improved documentation to special decommissioning tools and techniques - is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given

  18. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    1982-01-01

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

  19. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

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

  20. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  1. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  2. Light water reactor safeguards system evaluation

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

    1978-01-01

    A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

  3. Source-to-incident flux relation for a tokamak fusion test reactor blanket module

    International Nuclear Information System (INIS)

    Imel, G.R.

    1982-01-01

    The source-to-incident 14-MeV flux relation for a blanket module on the Tokamak Fusion Test Reactor is derived. It is shown that assumptions can be made that allow an analytical expression to be derived, using point kernel methods. In addition, the effect of a nonuniform source distribution is derived, again by relatively simple point kernel methods. It is thought that the methodology developed is valid for a variety of blanket modules on tokamak reactors

  4. The feature of high flux engineering test reactor and its role in nuclear power development

    International Nuclear Information System (INIS)

    Lu Guangquan

    1987-01-01

    The High Flux Engineering Test Reactor (HFETR) designed and built by Chinese own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11. 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR. (author)

  5. 1980 Annual status report: operation of the high flux reactor

    International Nuclear Information System (INIS)

    1981-01-01

    HFR Petten has been operated in 1980 in fulfilment of the 1980/83 JRC Programme Decision. Both reactor operation and utilization data have been met within a few percent of the goals set out in the annual working schedule, in support of a large variety of research programmes. Major improvements to experimental facilities have been introduced during the year and future modernization has been prepared

  6. High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

    2017-01-24

    Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

  7. Trends in light water reactor dosimetry programs

    International Nuclear Information System (INIS)

    Rahn, F.J.; Serpan, C.Z.; Fabry, A.; McElroy, W.N.; Grundl, J.A.; Debrue, J.

    1977-01-01

    Dosimetry programs and techniques play an essential role in the continued assurance of the safety and reliability of components of light water reactors. Primary concern focuses on the neutron irradiation embrittlement of reactor pressure vessels and methods by which the integrity of a pressure vessel can be predicted and monitored throughout its service life. Research in these areas requires a closely coordinated program which integrates the elements of the calculational and material sciences, the development of advanced dosimetric techniques and the use of benchmarks and validation of these methods. The paper reviews the status of the various international efforts in the dosimetry area

  8. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

  9. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

    1987-01-01

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  10. Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

    International Nuclear Information System (INIS)

    Schweizer, Fernando Lage Araújo

    2014-01-01

    The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

  11. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

    2006-01-01

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

  12. Production of Sn-117m in the BR2 high-flux reactor.

    Science.gov (United States)

    Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

    2009-01-01

    The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

  13. High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

    International Nuclear Information System (INIS)

    Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

    1993-01-01

    An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ''engineered safety features,'' which, along with the use of high temperature capable materials further enhance its safety characteristics

  14. Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D.; Villard, J. F.; Lyoussi, A. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Reynard-Carette, C. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France); Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y. [DEN Reactor Studies Dept., French Nuclear Energy and Alternative Energies Commission, CEA Cadarache, 13108 Saint Paul-Lez-Durance (France); Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J. [Laboratoire Chimie Provence LCP UMR 6264, Univ. of Provence, Centre St. Jerome, 13397 Marseille Cedex 20 (France)

    2011-07-01

    We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

  15. Combined analysis of neutron and photon flux measurements for the Jules Horowitz reactor core mapping

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Bignan, G.; Chauvin, J. P.; Gonnier, C.; Guimbal, P.; Malo, J. Y.; Carette, M.; Janulyte, A.; Merroun, O.; Brun, J.; Zerega, Y.; Andre, J.

    2011-01-01

    We study the combined analysis of nuclear measurements to improve the knowledge of the irradiation conditions in the experimental locations of the future Jules Horowitz Reactor (JHR). The goal of the present work is to measure more accurately neutron flux, photon flux and nuclear heating in the reactor. In a Material Testing Reactor (MTR), nuclear heating is a crucial parameter to design the experimental devices to be irradiated in harsh nuclear conditions. This parameter drives the temperature of the devices and of the samples. The numerical codes can predict this parameter but in-situ measurements are necessary to reach the expected accuracy. For this reason, one objective of the IN-CORE program [1] is to study the combined measurements of neutron and photon flux and their cross advanced interpretation. It should be reminded that both neutron and photon sensors are not totally selective as their signals are due to neutron and photon interactions. We intend to measure the neutron flux by three different kinds of sensors (Uranium Fission chamber, Plutonium Fission chamber and Self Powered Neutron Detector), the photon flux by two different sensors (Ionization chamber and Self Powered Gamma Detector) and the nuclear heating by two different ones (Differential calorimeter and Gamma Thermometer). For the same parameter, we expect that the use of different kinds of sensors will allow a better estimation of the aimed parameter by mixing different spectrum responses and different neutron and gamma contributions. An experimental test called CARMEN-1 is scheduled in OSIRIS reactor (CEA Saclay - France) at the end of 2011, with the goal to map irradiation locations in the reactor reflector to get a first validation of the analysis model. This article focuses on the sensor selection for CARMEN-1 experiment and to the way to link neutron and photon flux measurements in view to reduce their uncertainties but also to better assess the neutron and photon contributions to nuclear

  16. Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

    Science.gov (United States)

    Töre, Candan; Ortego, Pedro

    2005-01-01

    The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

  17. Environmentally assisted cracking in Light Water Reactors

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  18. Analysis of Neutron Flux Distribution in Rsg-Gas Reactor With U-Mo Fuels

    Directory of Open Access Journals (Sweden)

    Taswanda Taryo

    2004-01-01

    Full Text Available The use of U-Mo fuels in research reactors seems to be promising and, recently, world researchers have carried out these such activities actively. The National Nuclear Energy Agency (BATAN which owns RSG-GAS reactor available in Serpong Research Center for Atomic Energy should anticipate this trend. It is, therefore, this research work on the use of U-Mo fuels in RSG-GAS reactor should be carried out. The work was focused on the analysis of neutron flux distribution in the RSG-GAS reactor using different content of molybdenum in U-Mo fuels. To begin with, RSG-GAS reactor core model was developed and simulated into X, Y and Z dimensions. Cross section of materials based on the developed cells of standard and control fuels was then generated using WIMS-D5-B. The criticality calculations were finally carried out applying BATAN-2DIFF code. The results showed that the neutron flux distribution obtained in U-Mo-fuel-based RSG-GAS core is very similar to those achieved in the 300-gram sillicide-fuel-based RSG-GAS reactor core. Indeed, the utilization of the U-Mo RSG-GAS core can be very similar to that of the high-density sillicide reactor core and even could be better in the future.

  19. Partitioning Water Vapor and Carbon Dioxide Fluxes using Correlation Analysis

    Science.gov (United States)

    Scanlon, T. M.

    2008-12-01

    A variety of methods are currently available to partition water vapor fluxes (into components of transpiration and direct evaporation) and carbon dioxide fluxes (into components of photosynthesis and respiration), using chambers, isotopes, and regression modeling approaches. Here, a methodology is presented that accounts for correlations between high-frequency measurements of water vapor (q) and carbon dioxide (c) concentrations being influenced by their non-identical source-sink distributions and the relative magnitude of their constituent fluxes. Flux-variance similarity assumptions are applied separately to the stomatal and the non-stomatal exchange, and the flux components are identified by considering the q-c correlation. Water use efficiency for the vegetation, and how it varies with respect to vapor pressure deficit, is the only input needed for this approach that uses standard eddy covariance measurements. The method is demonstrated using data collected over a corn field throughout a growing season. In particular, the research focuses on the partitioning of the water flux with the aim of improving how direct evaporation is handled in soil-vegetation- atmosphere transfer models over the course of wetting and dry-down cycles.

  20. Towards intrinsically safe light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hannerz, K

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

  1. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

  2. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  3. Advanced light water reactor plant

    International Nuclear Information System (INIS)

    Giedraityte, Zivile

    2008-01-01

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

  4. Neutron flux shape effects in large fast reactor safety calculations

    International Nuclear Information System (INIS)

    Galati, A.; Loizzo, P.; Musco, A.

    1978-01-01

    Three classes of accidents in a large fast reactor were studied by the two-dimensional core dynamics code NADYP-2. A Modified version of the code, including a point kinetics module, allowed comparison between 2D and 0D power, reactivity and temperature histories. A strong shape effect was evidenced by these calculations in the boiling phase of LOF accidents as well as in the accident generated by control rod removal. Some future possibilities of by passing the consequences of this effect are indicated

  5. External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)

    International Nuclear Information System (INIS)

    Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H.

    1989-01-01

    The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10 -4 . In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

  6. 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)

  7. Ingested water equilibrates isotopically with the body water pool of a shorebird with unrivaled water fluxes

    NARCIS (Netherlands)

    Visser, G.H.; Dekinga, A; Achterkamp, B.; Piersma, T.

    We investigated the applicability of H-2 to measure the amount of body water (TBW) and water fluxes in relation to diet type and level of food intake in a mollusk-eating shorebird, the Red Knot (Calidris canutus). Six birds were exposed to eight experimental indoor conditions. Average fractional H-2

  8. AXIFLUX, Cosine Function Fit of Experimental Axial Flux in Cylindrical Reactor

    International Nuclear Information System (INIS)

    Holte, O.

    1980-01-01

    1 - Nature of physical problem solved: Calculates the parameters of the cosine function that will best fit data from axial flux distribution measurements in a cylindrical reactor. 2 - Method of solution: Steepest descent for the minimization. 3 - Restrictions on the complexity of the problem: Number of measured points less than 200

  9. The operating experience and incident analysis for High Flux Engineering Test Reactor

    International Nuclear Information System (INIS)

    Zhao Guang

    1999-01-01

    The paper describes the incidents analysis for High Flux Engineering test reactor (HFETR) and introduces operating experience. Some suggestion have been made to reduce the incidents of HFETR. It is necessary to adopt new improvements which enhance the safety and reliability of operation. (author)

  10. Two-field and drift-flux models with application to nuclear reactor safety

    International Nuclear Information System (INIS)

    Travis, J.R.

    1986-01-01

    The ideas of the two-field (6 equation model) and drift-flux (4 equation model) description of two-phase flows are presented. Several example calculations relating to reactor safety are discussed and comparisons of the numerical results and experimental data are shown to be in good agreement. 16 refs., 32 figs

  11. Visualization of neutron flux and power distributions in TRIGA Mark II reactor as an educational tool

    International Nuclear Information System (INIS)

    Snoj, Luka; Ravnik, Matjaz; Lengar, Igor

    2008-01-01

    Modern Monte Carlo computer codes (e.g. MCNP) for neutron transport allow calculation of detailed neutron flux and power distribution in complex geometries with resolution of ∼1 mm. Moreover they enable the calculation of individual particle tracks, scattering and absorption events. With the use of advanced software for 3D visualization (e.g. Amira, Voxler, etc.) one can create and present neutron flux and power distribution in a 'user friendly' way convenient for educational purposes. One can view axial, radial or any other spatial distribution of the neutron flux and power distribution in a nuclear reactor from various perspectives and in various modalities of presentation. By visualizing the distribution of scattering and absorption events and individual particle tracks one can visualize neutron transport parameters (mean free path, diffusion length, macroscopic cross section, up-scattering, thermalization, etc.) from elementary point of view. Most of the people remember better, if they visualize the processes. Therefore the representation of the reactor and neutron transport parameters is a convenient modern educational tool for the (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. The visualization of neutron flux and power distributions in Jozef Stefan Institute TRIGA Mark II research reactor is treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. (authors)

  12. Flux perturbation factor in cobalt samples for the reactor production of Co-60

    International Nuclear Information System (INIS)

    Curzio, O.A.

    1976-07-01

    Total flux perturbation factor (F) is experimentally determined for hollow cylinder cobalt samples irradiated in the RA-3 reactor. F factor is studied for different thicknesses of the material and the values are compared with those theoretically estimated by Dwork for a similar. (author) [es

  13. Measurements of thermal and fast neutron fluxes at the TRIGA reactor

    International Nuclear Information System (INIS)

    Zerdin, F.; Grabovsek, Z.; Klinc, T.; Solinc, H.

    1966-01-01

    Gold foils were placed at different positions in the TRIGA reactor core and in the experimental devices. Absolute values of the thermal neutron flux at these positions were obtained by coincidence method. Preliminary fast neutron spectrum was measured by threshold detector and by 'Li 6 sandwich' detector. A short description of the applied method and obtained measurements results are included [sl

  14. A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

    International Nuclear Information System (INIS)

    Pond, R.B.; Hanan, N.A.; Matos, J.E.

    1997-01-01

    A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

  15. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

  16. Heavy water upgrading system in the Fugen heavy water reactor

    International Nuclear Information System (INIS)

    Matsushita, T.; Susaki, S.

    1980-01-01

    The heavy water upgrading system, which is installed in the Fugen heavy water reactor (HWR) was designed to reuse degraded heavy water generated from the deuteration-dedeuteration of resin in the ion exchange column of the moderator purification system. The electrolysis method has been applied in this system on the basis of the predicted generation rate and concentration of degraded heavy water. The structural feature of the electrolytic cell is that it consists of dual cylindrical electrodes, instead of a diaphragm as in the case of conventional water electrolysis. 2 refs

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

    International Nuclear Information System (INIS)

    Marques, Andre L.F.; Todreas, Neil E.; Driscoll, Michael J.

    2000-01-01

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

  18. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  19. The chemistry of water reactor fuel

    International Nuclear Information System (INIS)

    Potter, P.E.

    1990-01-01

    In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

  20. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  1. Materials technologies of light water reactors

    International Nuclear Information System (INIS)

    Begley, R.

    1984-01-01

    Satisfactory materials performance is a key element in achieving reliable operation of light water reactors. Outstanding performance under rigorous operational conditions has been exhibited by pressure boundary components, core internals, fuel cladding, and other critical components of these systems. Corrosion and stress corrosion phenomena have, however, had an impact on plant availability, most notably relating to pipe cracking in BWR systems and steam generator corrosion in PWR systems. These experiences have stimulated extensive development activities by the nuclear industry in improved NDE techniques, investigation of corrosion phenomena, as well as improved materials and repair processes. This paper reviews key materials performance aspects of light water reactors with particular emphasis on the progress which has been made in modeling of corrosion phenomena, control of the plant operating environment, advanced material development, and application of sophisticated repair procedures. Implementation of this technology provides the basis for improved plant availability

  2. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    Adinolfi, R.; Noviello, L.

    1990-01-01

    The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

  3. WRAP: a water reactor analysis package

    International Nuclear Information System (INIS)

    Anderson, M.M.

    1977-06-01

    The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

  4. Advanced light water reactors for the nineties

    International Nuclear Information System (INIS)

    Ross, F.A.; Sugnet, W.R.

    1987-01-01

    The EPRI/Industry advanced light water reactor (ALWR) program and the US Department of Energy ALWR program are closely coordinated to meet the common objective which is the availability of improved and simplified light water reactor plants that may be ordered in the next decade to meet new or replacement capacity requirements. The EPRI/Industry objectives, program participants, and foreign participants, utility requirements document, its organization and content, small plant conceptual design program, the DOE ALWR program, design verification program, General Electric ABWR design features, Combustion Engineering system design, mid-size plant development, General Electric SBWR objectives, Westinghouse/Burns and Roe design objectives, construction improvement, and improved instrumentation and control are discussed in the paper

  5. High conversion heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyawaki, Yoshio; Wakabayashi, Toshio.

    1989-01-01

    In the present invention, fuel rods using uranium-plutonium oxide mixture fuels are arranged in a square lattice at the same pitch as that in light water cooled reactor and heavy water moderators are used. Accordingly, the volume ratio (Vm/Vf) between the moderator and the fuel can be, for example, of about 2. When heavy water is used for the moderator (coolant), since the moderating effect of heavy water is lower than that of light water, a high conversion ratio of not less than 0.8 can be obtained even if the fuel rod arrangement is equal to that of PWR (Vm/Vf about 2). Accordingly, it is possible to avoid problems caused by dense arrangement of fuel rods as in high conversion rate light water cooled reactors. That is, there are no more troubles in view of thermal hydrodynamic characteristics, re-flooding upon loss of coolant accident, etc., as well as the fuel production cost is not increased. (K.M.)

  6. Systematic assembly homogenization and local flux reconstruction for nodal method calculations of fast reactor power distributions

    International Nuclear Information System (INIS)

    Dorning, J.J.

    1991-01-01

    A simultaneous pin lattice cell and fuel bundle homogenization theory has been developed for use with nodal diffusion calculations of practical reactors. The theoretical development of the homogenization theory, which is based on multiple-scales asymptotic expansion methods carried out through fourth order in a small parameter, starts from the transport equation and systematically yields: a cell-homogenized bundled diffusion equation with self-consistent expressions for the cell-homogenized cross sections and diffusion tensor elements; and a bundle-homogenized global reactor diffusion equation with self-consistent expressions for the bundle-homogenized cross sections and diffusion tensor elements. The continuity of the angular flux at cell and bundle interfaces also systematically yields jump conditions for the scaler flux or so-called flux discontinuity factors on the cell and bundle interfaces in terms of the two adjacent cell or bundle eigenfunctions. The expressions required for the reconstruction of the angular flux or the 'de-homogenization' theory were obtained as an integral part of the development; hence the leading order transport theory angular flux is easily reconstructed throughout the reactor including the regions in the interior of the fuel bundles or computational nodes and in the interiors of the pin lattice cells. The theoretical development shows that the exact transport theory angular flux is obtained to first order from the whole-reactor nodal diffusion calculations, done using the homogenized nuclear data and discontinuity factors, is a product of three computed quantities: a ''cell shape function''; a ''bundle shape function''; and a ''global shape function''. 10 refs

  7. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  8. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  9. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  10. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    Fabrega, S.

    1979-01-01

    The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

  11. Development of neutron irradiation embrittlement correlation of reactor pressure vessel materials of light water reactors

    International Nuclear Information System (INIS)

    Soneda, Naoki; Dohi, Kenji; Nomoto, Akiyoshi; Nishida, Kenji; Ishino, Shiori

    2007-01-01

    A large amount of surveillance data of the RPV embrittlement of the Japanese light water reactors have been compiled since the current Japanese embrittlement correlation has been issued in 1991. Understanding on the mechanisms of the embrittlement has also been greatly improved based on both experimental and theoretical studies. CRIEPI and the Japanese electric power utilities have started research project to develop a new embrittlement correlation method, where extensive study of the microstructural analyses of the surveillance specimens irradiated in the Japanese commercial reactors has been conducted. The new findings obtained from the experimental study are that the formation of solute-atom clusters with little or no copper is responsible for the embrittlement in low-copper materials, and that the flux effect exists especially in high-copper materials and this is supported by the difference in the microstructure of the high-copper materials irradiated at different fluxes. Based on these new findings, a new embrittlement correlation method is formulated using rate equations. The new methods has higher prediction capability than the current Japanese embrittlement correlation in terms of smaller standard deviation as well as smaller mean value of the prediction error. (author)

  12. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    Stancavage, P.

    1991-01-01

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  13. Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

    International Nuclear Information System (INIS)

    Koch, R.; Findeisen, A.

    1986-04-01

    The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

  14. Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

    International Nuclear Information System (INIS)

    Mo, S.C.

    1991-01-01

    The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed

  15. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-02-01

    The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate those features that leave it open to public distrust and obstructive intervention. This redesign appears feasible within the realm of proven technology in those fields (fuels, materials, water chemistry, waste technology, etc.) in which extended operating experience is essential for confidence in system performance. A pressurized water reactor outline design developed to achieve the above goal is presented. The key feature is the design of the primary system extracting heat from the core so that the latter is protected from damage caused by any credible system failure or any destructive intervention from the outside by either violent means (up to and including nonnuclear warfare) or by mistaken or malicious use of the plant control systems. Such a design objective can be achieved by placing the entire primary circulation system in a large pressurized pool of cold water with a high boric acid content. Enough water is provided in the pool to allow core-decay-heat removal by evaporation for at least one week following any incident with no cooling systems operating. Subsequently it is assumed that a supply of further water (a few cubic meters per hour) from the outside can be arranged, even without the presence of the plant operating personnel

  16. Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Cook, David Howard

    2009-01-01

    A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities

  17. Reactor core of light water-cooled reactor

    International Nuclear Information System (INIS)

    Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

    1996-01-01

    In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

  18. Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson

    1996-01-01

    A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

  19. Neutron disadvantage factors in heavy water and light water reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1966-01-01

    A number od heavy water and light water reactor cells are analyzed in this paper by applying analytical methods of neutron thermalization. Calculations done according to the one-group Amouyal-Benoist method are included in addition. Computer codes for ZUSE Z-23 computer were written by applying both methods. The obtained results of disadvantage factors are then compared to results obtained by one-group P 3 approximation and by multigroup K7-THERMOS code [sr

  20. Bayesian statistics applied to neutron activation data for reactor flux spectrum analysis

    International Nuclear Information System (INIS)

    Chiesa, Davide; Previtali, Ezio; Sisti, Monica

    2014-01-01

    Highlights: • Bayesian statistics to analyze the neutron flux spectrum from activation data. • Rigorous statistical approach for accurate evaluation of the neutron flux groups. • Cross section and activation data uncertainties included for the problem solution. • Flexible methodology applied to analyze different nuclear reactor flux spectra. • The results are in good agreement with the MCNP simulations of neutron fluxes. - Abstract: In this paper, we present a statistical method, based on Bayesian statistics, to analyze the neutron flux spectrum from the activation data of different isotopes. The experimental data were acquired during a neutron activation experiment performed at the TRIGA Mark II reactor of Pavia University (Italy) in four irradiation positions characterized by different neutron spectra. In order to evaluate the neutron flux spectrum, subdivided in energy groups, a system of linear equations, containing the group effective cross sections and the activation rate data, has to be solved. However, since the system’s coefficients are experimental data affected by uncertainties, a rigorous statistical approach is fundamental for an accurate evaluation of the neutron flux groups. For this purpose, we applied the Bayesian statistical analysis, that allows to include the uncertainties of the coefficients and the a priori information about the neutron flux. A program for the analysis of Bayesian hierarchical models, based on Markov Chain Monte Carlo (MCMC) simulations, was used to define the problem statistical model and solve it. The first analysis involved the determination of the thermal, resonance-intermediate and fast flux components and the dependence of the results on the Prior distribution choice was investigated to confirm the reliability of the Bayesian analysis. After that, the main resonances of the activation cross sections were analyzed to implement multi-group models with finer energy subdivisions that would allow to determine the

  1. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    Leroy, C.; Bitsch, D.; Millot, J.P.

    1985-10-01

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  2. Neutron energy spectrum flux profile of Ghana's miniature neutron source reactor core

    International Nuclear Information System (INIS)

    Sogbadji, R.B.M.; Abrefah, R.G.; Ampomah-Amoako, E.; Agbemava, S.E.; Nyarko, B.J.B.

    2011-01-01

    Highlights: → The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was studied. → Using 20,484 energy grids, the thermal, slowing down and fast neutron energy regions were studied. - Abstract: The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) x 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) x 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) x 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939-3.7880) x 10 -08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) x 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 x 10 -01 MeV and 8.2680 x 10 -01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) x 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

  3. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2006-02-15

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation.

  4. Evaluation of critical heat flux performances for design strategy of new research reactor nuclear fuels

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Bang, In Cheol; Lee, Kwi Lim; Jeong, Yong Hoon

    2006-02-01

    The present project investigated stable burnout heat flux correlations applicable to research reactor operation conditions of low pressure, low temperature and high flow rate. In addition, in series of thermal limits important to safety of the reactor, ONB and OFI correlations also were investigated. There are some world CHF databases for tube-inside flow. In order to design a research reactor, DNB is final design limit factor and so the collection of the data or correlation are very important. The optimal core cooling capability can be done by considering neutronics, economical efficiency, materials limit together through engineering judgement based on DNB correlations. The project collected the materials and correlations applicable to research reactor conditions. The correlations give a fundamental base for analyzing thermal limit factors and will be used helpfully in review of regulatory body and designer for safety evaluation

  5. Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

    International Nuclear Information System (INIS)

    Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

    1990-01-01

    The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

  6. Maximum neutron flux in thermal reactors; Maksimum neutronskog fluksa kod termalnih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1968-07-01

    Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples.

  7. Critical heat fluxes in tubular fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    Subbotin, V.I.; Alekseev, G.V.; Peskov, O.L.

    1974-01-01

    The results of the experiments carried out show that with appropriate choice of tube, type and dimensions of intensifier the attainment of critical conditions at certain parameters is not accompanied by sharp or considerable increases in temperature of the heat removing surface. Increase in power to above critical under these conditions does not lead to considerable variation in temperature either. Thus, it appears possible to change from heat removal by steam-water mixture to convective heat removal by wet steam without manifestation of intolerable temperature conditions of the heating surface (Fig. 6). A change to convective heat removal by wet steam is possible at different levels of heat fluxes which depend during constant conditions at the inlet on tube length and the degree of the disturbing influence on the flow. This is especially important since in principle the possibility arises for developing a power reactor with tubular fuel elements, in which a once-through cycle with steam superheat involving no intermediate separation can be realised

  8. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

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

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  9. Good practices in heavy water reactor operation

    International Nuclear Information System (INIS)

    2010-06-01

    The value and importance of organizations in the nuclear industry engaged in the collection and analysis of operating experience and best practices has been clearly identified in various IAEA publications and exercises. Both facility safety and operational efficiency can benefit from such information sharing. Such sharing also benefits organizations engaged in the development of new nuclear power plants, as it provides information to assist in optimizing designs to deliver improved safety and power generation performance. In cooperation with Atomic Energy of Canada, Ltd, the IAEA organized the workshop on best practices in Heavy Water Reactor Operation in Toronto, Canada from 16 to 19 September 2008, to assist interested Member States in sharing best practices and to provide a forum for the exchange of information among participating nuclear professionals. This workshop was organized under Technical Cooperation Project INT/4/141, on Status and Prospects of Development for and Applications of Innovative Reactor Concepts for Developing Countries. The workshop participants were experts actively engaged in various aspects of heavy water reactor operation. Participants presented information on activities and practices deemed by them to be best practices in a particular area for consideration by the workshop participants. Presentations by the participants covered a broad range of operational practices, including regulatory aspects, the reduction of occupational dose, performance improvements, and reducing operating and maintenance costs. This publication summarizes the material presented at the workshop, and includes session summaries prepared by the chair of each session and papers submitted by the presenters

  10. Light-water reactor research and development

    International Nuclear Information System (INIS)

    1985-05-01

    This report on the national program of research and development on light water reactors is the second of two reports requested in 1982 by W. Kenneth Davis, Deputy Secretary of the Department of Energy. A first report, published in September 1983, treated the needs for safety-related R and D. In this second report, the Energy Research Advisory Board finds that, although many light water reactors are providing reliable and economic electricity, it appears unlikely that U.S. utilities will order additional reactors until the currently unacceptable economic risk, created by the regulatory climate and uncertain demand, is reduced. Thus it is unlikely that the private sector alone will fund major LWR design improvements. However, nuclear power will continue on its current course of expansion overseas. DOE participation is vitally needed to support the national interest in LWR technology. The report outlines R and D needs for a program to improve the safety, reliability, and economics of the present generation of plants; to develop evolutionary improved designs to be ready when needed; and to explore innovative longer-term concepts for deployment after the year 2000. The respective roles of government and the private sector are discussed

  11. CFD validation of a supercritical water flow for SCWR design heat and mass fluxes

    International Nuclear Information System (INIS)

    Roelofs, F.; Lycklama a Nijeholt, J.A.; Komen, E.M.J.; Lowenberg, M.; Starflinger, J.

    2007-01-01

    The applicability of Computational Fluid Dynamics (CFD) for water under supercritical conditions in supercritical water reactors (SCWR) has still to be verified. In the recent past, CFD validation analyses were performed by various institutes for supercritical water in vertical tubes based on the well known experimental data from Yamagata. However, validation using data from experiments with working conditions closer to the actual operational conditions of such reactors is needed. From a literature survey the experiments performed by Herkenrath are selected to perform validation analyses at higher heat fluxes and a higher mass flux. The accuracy of CFD using RANS (Reynolds Average Navier-Stokes) turbulence modelling for supercritical fluids under conditions close to the operational conditions of a supercritical water reactor is determined. It is concluded that the wall temperature can be predicted by RANS CFD, using the RNG k-ε turbulence model, with accuracy in the range of 5% for heat fluxes up to 1100 kW/m 2 and for a bulk enthalpy up to 2200 kJ/kg. For a bulk enthalpy exceeding 2200 kJ/kg, a significant lower accuracy of the CFD predictions (about 3%) is found for the simulations of the experiments of Yamagata in comparison with the simulations of the experiments of Herkenrath. For these experiments, the accuracy is about 18 per cent. This might be a result of the fact that the CFD analyses do not simulate the flattening of the temperature profile at about 2200 kJ/kg which is found in the experiments of Herkenrath. However, the obtained accuracies ranging from 3% to 18% are still deemed to be acceptable for many design purposes. (authors)

  12. Water reactor fuel research at the Joint Research Centre of the Commission of the European Union

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J [HFR Unit, Inst. for Advanced Materials, Petten (Netherlands)

    1997-12-01

    The JRC programmes 1995-1998 are in progress within 8 JRC Institutes located within different research centres situated in five member states of the European Union. Except for the area of reactor safety and waste management, there are no JRC research programmes related to water reactor fuel directly. However, the JRC is providing support on basis of contracts and cooperations to bilateral and international R and D programmes through access to its large test facilities like the High Flux Reactor at the JRC Petten and the hot cell facilities of the Transuranium Institute at the JRC Karlsruhe. Access is available to customers from all over the world. 3 tabs.

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

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

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

  15. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  16. Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

  17. Controlling hydrogen behavior in light water reactors

    International Nuclear Information System (INIS)

    Cullingford, H.S.; Edeskuty, F.J.

    1981-01-01

    In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

  18. Status of advanced small pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Peipei; Zhou Yun

    2012-01-01

    In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

  19. On the determination of boiling water reactor characteristics by noise analysis

    International Nuclear Information System (INIS)

    Kleiss, J.

    1983-01-01

    In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

  20. Neutron Flux Variation in the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Yahaya, M.; Ahmed, Y.A.

    2013-01-01

    In order to ascertain the level of flux variation in one of the inner irradiation channels of the Nigeria Research Reactor-1 (NIRR-1), the irradiation container used for routine activation analysis was employed with copper wires as flux monitors. Measurements were carried out with copper wires arranged in axial direction to determine the thermal neutron flux at selected positions using absolute foil activation method. Our results show that there exists a slight flux variation from one position to another ranging from (4.57±0.21) x 10 11 to (5.20± 0.20) x 10 11 cm -2 s -1 .Individual foil shows slight flux variation from one position to another in the same irradiation container but they all pointed toward a level of consistency in variation in spite of the recent installation of the cadmium lined irradiation channel. The values obtained in this work are in good agreement with the previously measured value of (5.14±0.24) x 10 11 cm -2 s -1 after commissioning of NIRR-1 (Jonah et al., 2005). This shows that the cadmium lined installation does not affect the flux stability. In order to improve the accuracy of NAA using NIRR-l facility, there is need for flux corrections to be made by MNSR users during NAA particularly for samples in the axial position for long irradiation.

  1. Assessment of gold flux monitor at irradiation facilities of MINT TRIGA MK II reactor

    International Nuclear Information System (INIS)

    Wee Boon Siong; Abdul Khalik Wood; Mohd Suhaimi Hamzah; Shamsiah Abdul Rahman; Md Suhaimi Elias; Nazaratul Ashifa Abd Salim

    2005-01-01

    Neutron source of MINTs TRIGA MK II reactor has been used for activation analysis for many years and neutron flux plays important role in activation of samples at various positions. Currently, two irradiation facilities namely the pneumatic transfer system and rotary rack are available to cater for short and long lived irradiation. Neutron flux variation for both irradiation facilities have been determined using gold wire and gold solution as flux monitor. However, the use of gold wire as flux monitor is costlier if compared to gold solution. The results from analysis of certified reference materials showed that gold solution as flux monitors yield satisfactory results and proved to safe cost on the purchasing of gold wire. Further experiment on self-shielding effects of gold solution at various concentrations has been carried out. This study is crucial in providing vital information on the suitable concentration for gold solution as flux monitor. In the near future, gold solution flux monitor will be applied for routine analysis and hence to improve the capability of the laboratory on neutron activation analysis. (Author)

  2. Scram and nonlinear reactor system seismic analysis for the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Morrone, A.

    1975-01-01

    A description is given of the analysis and results for the Fast Flux Test Facility (FFTF) reactor system which was analyzed for both scram times and seismic responses such as bending moments and impact forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node. The results give time history plots of various seismic responses and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. (U.S.)

  3. Fast flux measurements by means of threshold detectors on the reactor 'Melusine'

    International Nuclear Information System (INIS)

    Leger, P.; Sautiez, B.

    1959-01-01

    Using existing data on the (n,p) and (n,α) threshold reactions we have carried out fast flux measurements on the swimming pool type reactor 'Melusine'. Four common elements: P, S, Mg, Al were chosen because from the point of view of fast spectrum analysis they represent a fairly good energy range from 2.4 MeV to 8 MeV. The fission flux value found in the central element at a power of 1 MW is 1.4 x 10 13 n/cm 2 /s ± 0.14. (author) [fr

  4. Absolute measurements of the fast neutron flux in the reactor RA

    Energy Technology Data Exchange (ETDEWEB)

    Berovic, N; Boreli, F; Dragin, R [Institute of Nuclear Sciences Boris Kidric, Department of physics, Vinca, Beograd (Serbia and Montenegro)

    1961-10-15

    The absolute neutron flux in the vertical VK-5 hole of the reactor RA was determined by using the {sup 27}Al (n, alpha) {sup 24}Na reaction, and by counting the {sup 24}Na - 2.5 MeV gamma line photopeak activity. A method for the determination of {sigma}{sub eff} as a mean value between the two large limiting cases of neutron spectra is used. The flux at the power level of 5 MW was found to be (2.5{+-}0.9){center_dot}10{sup 12}n/cm{sup 2}sec (author)

  5. An overview of FFTF [Fast Flux Test Facility] contributions to Liquid Metal Reactor Safety

    International Nuclear Information System (INIS)

    Waltar, A.E.; Padilla, A. Jr.

    1990-11-01

    The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs

  6. Determination of fast neutron flux distribution in irradiation sites of the Malaysian Nuclear Agency research reactor

    International Nuclear Information System (INIS)

    Yavar, A.R.; Sarmani, S.B.; Wood, A.K.; Fadzil, S.M.; Radir, M.H.; Khoo, K.S.

    2011-01-01

    Determination of thermal to fast neutron flux ratio (f fast ) and fast neutron flux (φ fast ) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f fast and subsequently φ fast were determined using the absolute method. The f fast ranged from 48 to 155, and the φ fast was found in the range 1.03x10 10 -4.89x10 10 n cm -2 s -1 . These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.

  7. Water-fluxed melting of the continental crust: A review

    Czech Academy of Sciences Publication Activity Database

    Weinberg, R. F.; Hasalová, Pavlína

    212-215, January (2015), s. 158-188 ISSN 0024-4937 Institutional support: RVO:67985530 Keywords : aqueous fluids * crustal anatexis * granites * silicate melts * water-fluxed melting Subject RIV: DB - Geology ; Mineralogy Impact factor: 3.723, year: 2015

  8. Precipitable water and vapor flux between Belem and Manaus

    International Nuclear Information System (INIS)

    Marques, J.

    1977-01-01

    The water vapor flux and precipitable water was computated over the natural Amazon forest in the stretch between Belem and Manaus for 1972. The atmospheric branch of hidrological cycle theory was applied and the most significant conclusions on an annual basis are: Atlantic Ocean water vapor contributes 52% to the regional precipitation and is significant the role played by local evapotranspiration in the precipitation in the area; there were signs of the phenomenon of water vapor recycling nearly throughout the year. Evapotranspiration contributes to 48% of the precipitations in the area studied. The real evapotranspiration estimated by this method was 1,000mm year - 1 [pt

  9. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  10. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  11. Control of water chemistry in operating reactors

    International Nuclear Information System (INIS)

    Riess, R.

    1997-01-01

    Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ''modified'' B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs

  12. Control of water chemistry in operating reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riess, R [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ``modified`` B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs.

  13. 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

  14. Measurement of the epithermal neutron flux of the Argonauta reactor by the Sandwich method

    International Nuclear Information System (INIS)

    Nascimento, H.M.

    1973-01-01

    A common method of obtaining information about the neutron spectrum in the energy range of 1 eV to a few keV is by using resonance sandwich detectors. A sandwich detector is usually made up of three foils placed one on top of the other, each having the same thickness and being made of the same material which has a pronounced absorption resonance. To make an adequate evaluation, the sandwich method was compared with one using an isolated detector. The results obtained from approximate theoretical calculations were checked experimentally, using In, Au and Mn foils, in an isotropic 1/E flux in the Argonaut Reactor at I.E.N. As practical application of this method, the deviation from a 1/E spectrum of the epithermal neutron flux in the core and external graphite reflector of the Argonaut Reactor has been measured with the sandwich foils previously calibrated in a 1/E spectrum. (author)

  15. An optimization study of peak thermal neutron flux in moderators of advanced repetitive pulse reactors

    International Nuclear Information System (INIS)

    Asaoka, Takumi; Watanabe, N.

    1976-01-01

    In achieving a high peak thermal neutron flux in hydrogenous moderators installed in repetitive pulse reactors, the core-moderator arrangement can play as much an important role as the moderator design itself. However, the effect of the former has not been adequately emphasized to date, while a rather extensive study has been made on the latter. The present study concerns with a core-moderator system parameter optimization for a repetitive accelerator pulsed fast reactor. The results have shown that small differences in the arrangement resulting from the optimizations of various parameters are significant and the effects can be summed up to give an increase in the peak thermal flux by a factor of about two. (auth.)

  16. The interpretation of neutron noise in boiling water reactors

    International Nuclear Information System (INIS)

    John, T.M.; Singh, O.P.

    1985-01-01

    Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

  17. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  18. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  19. Is light water reactor technology sustainable?

    International Nuclear Information System (INIS)

    Rothwell, G.; Van der Zwaan, B.

    2001-01-01

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

  20. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  1. Westinghouse Water Reactor Divisions quality assurance plan

    International Nuclear Information System (INIS)

    1977-09-01

    The Quality Assurance Program used by Westinghouse Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements. This program satisfies the NRC Quality Assurance Criteria, 10CFR50 Appendix B, to the extent that these criteria apply to safety related NSSS equipment. Also, it follows the regulatory position provided in NRC regulatory guides and the requirements of ANSI Standard N45.2.12 as identified in this Topical Report

  2. Light-water-reactor hydrogen manual

    International Nuclear Information System (INIS)

    Camp, A.L.; Cummings, J.C.; Sherman, M.P.; Kupiec, C.F.; Healy, R.J.; Caplan, J.S.; Sandhop, J.R.; Saunders, J.H.

    1983-06-01

    A manual concerning the behavior of hydrogen in light water reactors has been prepared. Both normal operations and accident situations are addressed. Topics considered include hydrogen generation, transport and mixing, detection, and combustion, and mitigation. Basic physical and chemical phenomena are described, and plant-specific examples are provided where appropriate. A wide variety of readers, including operators, designers, and NRC staff, will find parts of this manual useful. Different sections are written at different levels, according to the most likely audience. The manual is not intended to provide specific plant procedures, but rather, to provide general guidance that may assist in the development of such procedures

  3. Development trends in light water reactors

    International Nuclear Information System (INIS)

    Fogelstroem, L.; Simon, M.

    1988-01-01

    The present market for new nuclear power plants is weak, but is expected to pick up again, which is why great efforts are being made to further develop the light water reactor line for future applications. There is both a potential and a need for further improvement, for instance with respect to even higher cost efficiency, a simplified operating permit procedure, shorter construction periods, and increased operational flexibility to meet rising demands in load following behavior and in better cycle data of fuel elements. However, also public acceptance must not be forgotten when deciding about the line to be followed in the development of LWR technology. (orig.) [de

  4. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    Yoshio Matsuo

    1987-01-01

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

  5. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  6. Thermodynamic analysis of a supercritical water reactor

    International Nuclear Information System (INIS)

    Edwards, M.

    2007-01-01

    A thermodynamic model has been developed for a hypothetical design of a Supercritical Water Reactor, with emphasis on Canadian design criteria. The model solves for cycle efficiency, mass flows and physical conditions throughout the plant based on input parameters of operating pressures and efficiencies of components. The model includes eight feedwater heaters, three feedwater pumps, a deaerator, a condenser, the core, three turbines and two reheaters. To perform the calculations, Microsoft Excel was used in conjunction with FLUIDCAL-IAPWS95 and VBA code. The calculations show that a thermal efficiency of 47.5% can be achieved with a core outlet temperature of 625 o C. (author)

  7. An analog computer method for solving flux distribution problems in multi region nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Radanovic, L; Bingulac, S; Lazarevic, B; Matausek, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1963-04-15

    The paper describes a method developed for determining criticality conditions and plotting flux distribution curves in multi region nuclear reactors on a standard analog computer. The method, which is based on the one-dimensional two group treatment, avoids iterative procedures normally used for boundary value problems and is practically insensitive to errors in initial conditions. The amount of analog equipment required is reduced to a minimum and is independent of the number of core regions and reflectors. (author)

  8. Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

    International Nuclear Information System (INIS)

    Engle, W.W. Jr.; Williams, L.R.

    1994-07-01

    This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

  9. Seismic, high wind, tornado, and probabilistic risk assessment of the high flux isotope reactor

    International Nuclear Information System (INIS)

    Harris, S.P.; Hashimoto, P.S.; Dizon, J.O.; Hashimoto, P.S.

    1989-01-01

    Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR). Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed

  10. Seismic, high wind, tornado, and probabilistic risk assessments of the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Harris, S.P.; Stover, R.L.; Hashimoto, P.S.; Dizon, J.O.

    1989-01-01

    Natural phenomena analyses were performed on the High Flux Isotope Reactor (HFIR) Deterministic and probabilistic evaluations were made to determine the risks resulting from earthquakes, high winds, and tornadoes. Analytic methods in conjunction with field evaluations and an earthquake experience data base evaluation methods were used to provide more realistic results in a shorter amount of time. Plant modifications completed in preparation for HFIR restart and potential future enhancements are discussed. 5 figs

  11. Measurement of thermal, epithermal and fast neutron flux in the IEA-R1 reactor by the foil activation method

    International Nuclear Information System (INIS)

    Koskinas, M.F.

    1979-01-01

    Experimental and theoretical details of the foil activation method applied to neutrons flux measurements at the IEA-R1 reactor are presented. The thermal - and epithermal - neutron flux were determined form activation measurements of gold, cobalt and manganese foils; and for the fast neutron flux determination, aluminum, iron and nickel foils were used. The measurements of the activity induced in the metal foils were performed using a Ge-Li gamma spectrometry system. In each energy range of the reactor neutron spectrum, the agreement among the experimental flux values obtained using the three kind of materials, indicates the consistency of the theoretical approach and of the nuclear parameters selected. (Author) [pt

  12. Production of high-specific activity radionuclides using SM high-flux reactor

    International Nuclear Information System (INIS)

    Karelin, Ye.A.; Toporov, Yu.G.; Filimonov, V.T.; Vakhetov, F.Z.; Tarasov, V.A.; Kuznetsov, R.A.; Lebedev, V.M.; Andreev, O.I.; Melnik, M.I.; Gavrilov, V.D.

    1997-01-01

    The development of High Specific Activity (HSA) radionuclides production technologies is one of the directions of RIAR activity, and the high flux research reactor SM, having neutron flux density up to 2.10 15 cm -2 s 1 in a wide range of neutron spectra hardness, plays the principal role in this development. The use of a high-flux reactor for radionuclide production provides the following advantages: production of radionuclides with extremely high specific activity, decreasing of impurities content in irradiated targets (both radioactive and non-radioactive), cost-effective use of expensive isotopically enriched target materials. The production technologies of P-33, Gd-153, W-188, Ni-63, Fe-55,59, Sn-113,117m,119m, Sr- 89, applied in industry, nuclear medicine, research, etc, were developed by RIAR during the last 5-10 years. The research work included the development of calculation procedures for radionuclide reactor accumulation forecast, experimental determination of neutron cross-sections, the development of irradiated materials reprocessing procedures, isolation and purification of radionuclides. The principal results are reviewed in the paper. (authors)

  13. Management of safety and risk at the HFIR [High-Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Glovier, H.A.

    1990-01-01

    This paper discusses the management of safety and risk at the High-Flux Isotope Reactor (HFIR), a category A research reactor at Oak Ridge National Laboratory (ORNL). The HFIR went critical in 1966 and operated at its designed 100 MW for 20 yr until it was shut down on November 14, 1986. It operated at a >90% availability and without significant event during this period. The result was a complacent management program lacking rigor. This complacency came to an end with the Chernobyl accident, which led to the appointment of an internal committee to assess the safety of ORNL reactor operations. This committee found that HFIR pressure vessel material specimens removed several years earlier had not been analyzed. This issue led to a general review of management practices that were found lacking in quality assurance, safety documentation, training process, and emergency planning, among others. Management accountability was lacking, as shown by design basis and safety analyses that were not up to data and by the fact that reactor operators whose requalification examinations had not been graded were allowed to continue operating the reactor over a long period of time. Between shutdown in 1986 and restart in April 1989, significant management changes and initiatives were made in the area of risk and safety management of ORNL reactors. These are presented briefly in this paper

  14. Method and apparatus for controlling the neutron flux in nuclear reactors

    International Nuclear Information System (INIS)

    Minnick, L.E.

    1979-01-01

    A control rod assembly in a nuclear reactor that automatically scrams the reactor when a loss of coolant flow occurs and that can also control the level of neutron flux in the reactor is described. The control rod assembly includes a separator plate having an orifice through which the reactor coolant flows and a sealing surface around the orifice. The control rod in the assembly has a complementary sealing surface. When the control rod and separator plate are brought into contact, the differential pressure across the separator plate caused by the flow of the primary coolant through the reactor core retains the two sealing surfaces together. If the flow of coolant stops or the differential pressure across the separator plate decreases for any reason, the control rod drops by gravity and the reactor is scrammed. The control rod is also automatically dropped as a result of the lateral vibration of an earthquake or by the downward motion of the rod drive shaft, either of which will open the sealing surfaces and reduce the sealing pressure

  15. A new mechanism of hydrogen absorption in water-water reactor core materials

    International Nuclear Information System (INIS)

    Gann, V.V.; Gann, A.V.

    2012-01-01

    The spectrum of fast protons, generated in water by fast neutrons of WWER-1000 reactor core, has been calculated using the code MCNPX. The main mechanism of fast proton generation in the moderator is found to be elastic scattering of fast neutrons on hydrogen nuclei. Fast protons with mean energy 1 MeV flow towards the surface of cladding material at flux density ∼ 0.1 μA/cm 2 . Proton range distribution profile in cladding material is calculated. The range of fast protons in zirconium averages 20 μm, the maximal proton range is larger than 200 μm. The rate of hydrogen deposition in 40 μm layer amounts to 5 x 10 -5 H/n/μ. A role of the suggested mechanism in process of zirconium clad hydrogenation during reactor irradiation is discussed.

  16. A new detector for the measurement of neutron flux in nuclear reactors

    International Nuclear Information System (INIS)

    Koch, L.; Labeyrie, J.; Tarassenko, S.

    1958-01-01

    The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high γ ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their β and γ disintegration. Thermal neutron fluxes as low as few neutrons per cm 2 per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm 2 ) for fluxes higher than 10 8 n/cm 2 /s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author) [fr

  17. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P. C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  18. Screening reactor steam/water piping systems for water hammer

    International Nuclear Information System (INIS)

    Griffith, P.

    1997-09-01

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

  19. Radiological considerations of the reactor cover gas processing system at the FFTF [Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Prevo, P.R.

    1986-09-01

    Radiological and environmental protection experience associated with the reactor cover gas processing system at the Fast Flux Test Facility (FFTF) has been excellent. Personnel radiation exposures received from operating and maintaining the reactor cover gas processing system have been very low, the system has remained free of radioactive particulate contamination through the first seven operating cycles (cesium contamination was detected at the end of Cycle 8A), and releases of radioactivity to the environment have been very low, well below environmental standards. This report discusses these three aspects of fast reactor cover gas purification over the first eight operating cycles of the FFTF (a duration of a little more than four years, from April 1982 through July 1986)

  20. The proposed use of low enriched uranium fuel in the High Flux Australian Reactor (HIFAR)

    International Nuclear Information System (INIS)

    Vittorio, D.; Durance, G.

    2002-01-01

    The Australian Nuclear Science and Technology Organisation (ANSTO) operates the High Flux Australian Reactor (HIFAR). HIFAR commenced operation in the late 1950's with fuel elements containing uranium enriched to 93%. From that time the level of enrichment has gradually decreased to the current level of 60%. It is now proposed to further reduce the enrichment of HIFAR fuel to <20% by utilising LEU fuel assemblies manufactured by RISO National Laboratory, that were originally intended for use in the DR-3 reactor. Minor modifications have been made to the assemblies to adapt them for use in HIFAR. A detailed design review has been performed and initial safety analysis and reactor physics calculations are to be submitted to ARPANSA as part of a four-stage approval process. (author)

  1. Neutron flux measurement in the thermal column of the Malaysian TRIGA mark II reactor with MCNP verification

    International Nuclear Information System (INIS)

    Abdel Munem, E.; Shukri, A.; Tajuddin, A.A.

    2006-01-01

    A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)

  2. Evaluation of HFIR [High Flux Isotope Reactor] pressure-vessel integrity considering radiation embrittlement

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Merkle, J.G.; Nanstad, R.K.

    1988-04-01

    The High Flux Isotope Reactor (HFIR) pressure vessel has been in service for 20 years, and during this time, radiation damage was monitored with a vessel-material surveillance program. In mid-November 1986, data from this program indicated that the radiation-induced reduction in fracture toughness was greater than expected. As a result, a reevaluation of vessel integrity was undertaken. Updated methods of fracture-mechanics analysis were applied, and an accelerated irradiations program was conducted using the Oak Ridge Research Reactor. Results of these efforts indicate that (1) the vessel life can be extended 10 years if the reactor power level is reduced 15% and if the vessel is subjected to a hydrostatic proof test each year; (2) during the 10-year life extension, significant radiation damage will be limited to a rather small area around the beam tubes; and (3) the greater-than-expected damage rate is the result of the very low neutron flux in the HFIR vessel relative to that in samples of material irradiated in materials-testing reactors (a factor of ∼10 4 less), that is, a rate effect

  3. Development, Implementation and Experimental Validations of Activation Products Models for Water Pool Reactors

    International Nuclear Information System (INIS)

    Petriw, S.N.

    2001-01-01

    Some parameters were obtained both calculations and experiments in order to determined the source of the meaning activation products in water pool reactors. In this case, the study was done in RA-6 reactor (Centro Atomico Bariloche - Argentina).In normal operation, neutron flux on core activates aluminium plates.The activity on coolant water came from its impurities activation and meanly from some quantity of aluminium that, once activated, leave the cladding and is transported by water cooling system.This quantity depends of the 'recoil range' of each activation reaction.The 'staying time' on pool (the time that nuclides are circulating on the reactor pool) is another characteristic parameter of the system.Stationary state activity of some nuclides depends of this time.Also, several theoretical models of activation on coolant water system are showed, and their experimental validations

  4. Coarse mesh finite element method for boiling water reactor physics analysis

    International Nuclear Information System (INIS)

    Ellison, P.G.

    1983-01-01

    A coarse mesh method is formulated for the solution of Boiling Water Reactor physics problems using two group diffusion theory. No fuel assembly cross-section homogenization is required; water gaps, control blades and fuel pins of varying enrichments are treated explicitly. The method combines constrained finite element discretization with infinite lattice super cell trial functions to obtain coarse mesh solutions for which the only approximations are along the boundaries between fuel assemblies. The method is applied to bench mark Boiling Water Reactor problems to obtain both the eigenvalue and detailed flux distributions. The solutions to these problems indicate the method is useful in predicting detailed power distributions and eigenvalues for Boiling Water Reactor physics problems

  5. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Boåsen, Magnus, E-mail: boasen@kth.se [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Ehrnstén, Ulla [VTT Technical Research Centre of Finland Ltd, PO Box 1000, FI-02044 VTT (Finland)

    2017-02-15

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations. - Highlights: • Hardness testing is combined with post irradiation annealing at 330, 360 and 390 °C. • Unstable matrix defects is studied in a reactor pressure vessel steel. • Comparison between surveillance material and accelerated irradiation. • No evidence of unstable matrix defects, i.e. not present in studied material. • Difference in hardness recovery between irradiation conditions found at 390 °C.

  6. Measurements of neutron flux distributions in the core of the Ljubljana TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    Rant, J.; Ravnik, M.; Mele, I.; Dimic, V.

    2008-01-01

    Recently the Ljubljana TRIGA Mark II Reactor has been refurbished and upgraded to pulsed operation. To verify the core design calculations using TRIGAP and PULSTR1 codes and to obtain necessary data for future irradiation and neutron beam experiments, an extensive experimental program of neutron flux mapping and neutron field characterization was carried out. Using the existing neutron measuring thimbles complete axial and radial distributions in two radial directions were determined for two different core configurations. For one core configuration the measurements were also carried out in the pulsed mode. For flux distributions thin Cu (relative measurements) and diluted Au wires (absolute values) were used. For each radial position the cadmium ratio was determined in two axial levels. The core configuration was rather uniform, well defined (fresh fuel of a single type, including fuelled followers) and compact (no irradiation channels or gaps), offering unique opportunity to test the computer codes for TRIGA reactor calculations. The neutron flux measuring procedures and techniques are described and the experimental results are presented. The agreement between the predicted and measured power peaking factors are within the error limits of the measurements (<±5%) and calculations (±10%). Power peaking occurs in the B ring, and in the A ring (centre) there is a significant flux depression. (authors)

  7. Development of a 10-decade single-mode reactor flux monitoring system

    International Nuclear Information System (INIS)

    Valentine, K.H.; Shepard, R.L.; Falter, K.G.; Reese, W.B.

    1988-01-01

    Conventional wide-range neutron channels employ three optional modes to monitor the required flux range from source levels to full power (typically 10 or more decades). Difficult calibrations are necessary to provide a continuous output signal when such a system switches from counting mode in the source range to mean-square voltage mode in the midrange to dc current mode in the power range. In an ORNL proof-of-principle test, a method of extended range counting was implemented with a fission counter and conventional wide-band pulse processing electronics to provide a single-mode, monotonically increasing signal that spanned /approximately 10/ decades of neutron flux. Ongoing work includes design, fabrication, and testing of a comlpete neutron flux monitoring system suitable for advanced liquid metal reactor designs. 6 refs., 4 figs

  8. Analysis of the Daya Bay Reactor Antineutrino Flux Changes with Fuel Burnup

    Science.gov (United States)

    Hayes, A. C.; Jungman, Gerard; McCutchan, E. A.; Sonzogni, A. A.; Garvey, G. T.; Wang, X. B.

    2018-01-01

    We investigate the recent Daya Bay results on the changes in the antineutrino flux and spectrum with the burnup of the reactor fuel. We find that the discrepancy between current model predictions and the Daya Bay results can be traced to the original measured U 235 /Pu 239 ratio of the fission β spectra that were used as a base for the expected antineutrino fluxes. An analysis of the antineutrino spectra that is based on a summation over all fission fragment β decays, using nuclear database input, explains all of the features seen in the Daya Bay evolution data. However, this summation method still allows for an anomaly. We conclude that there is currently not enough information to use the antineutrino flux changes to rule out the possible existence of sterile neutrinos.

  9. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

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

    1996-07-01

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

  10. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

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

  11. Carbon-14 in reactor plant water

    International Nuclear Information System (INIS)

    Knowles, G.K.

    1979-01-01

    The method for the analysis of 14 C in reactor plant water and various waste streams previously used at the Idaho National Engineering Laboratory has been shown to be ineffective for samples which contain organic compounds. The previous method consisted of acidification and refluxing of the sample, precipitation of the liberated CO 2 , and subsequent analysis by the liquid scintillation method. The method was simple but it did not convert all compounds containing 14 C in the sample to CO 2 . The new method, while it is based on the previous method, has been improved by employing a strong oxidant, potassium persulfate and silver nitrate, for more complete oxidation of the organics to CO 2 . The new method yields 14 C values that have typically been one to two orders of magnitude higher than the values obtained using the former method. This indicates that most of the 14 C present in the current reactor water samples being analyzed is associated with trace amounts of organics

  12. Thermal neutron flux distribution in the ET R R-1 reactor core as experimentally measured and theoretically calculated by the code triton

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National center for nuclear safety and radiation control, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    Thermal neutron flux distributions that were measured earlier at the ET-R R-1 reactor are compared with those calculated by the three dimensional diffusion code Triton. This comparison was made for the horizontal and vertical flux distributions. The horizontal thermal flux distributions considered in this comparison were along the core diagonals at two planes of different heights from core bottom, where one at a level passing through the control rod at core center and the other at a level below this control rod. In the meantime all the control rods were taken into consideration. The effect of the existence of a water cavity inside the core as well as the influence of the control rods on the thermal flux are illustrated in this work. The vertical thermal flux distributions considered in the comparison were at two positions in core namely; one along the core height the horizontal reactor power distribution along the core height and the horizontal reactor power distribution along the core diagonal as calculated by the code Triton are also given this work. 8 figs., 1 tab.

  13. Neutronic study of the two french heavy water reactors

    International Nuclear Information System (INIS)

    Horowitz, J.

    1955-01-01

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

  14. Water inventory management in condenser pool of boiling water reactor

    International Nuclear Information System (INIS)

    Gluntz, D.M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs

  15. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

    International Nuclear Information System (INIS)

    Merz, Stefan; Djuricic, Mile; Villa, Mario; Boeck, Helmuth; Steinhauser, Georg

    2011-01-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10 9 cm -2 s -1 at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. - Highlights: → Neutron activation is an important process for the waste management of nuclear facilities. → Biological shield of the TRIGA reactor Vienna has been topic of investigation. → Flux values allow a categorization of the concrete concerning radiation protection legislation. → Reactor installations are of great importance as neutron sources into the biological shield. → Every installation shows distinguishable flux profiles.

  16. NEPTUNE: a modular system for light-water reactor calculation

    International Nuclear Information System (INIS)

    Bouchard, J.; Kanevoky, A.; Reuss, P.

    1975-01-01

    A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

  17. Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

    International Nuclear Information System (INIS)

    Khattab, K.; Omar, H.; Ghazi, N.

    2009-01-01

    A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

  18. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  19. Measure of thermal neutron flux in the IPEN/MB-01 reactor using 197 Au wire activation detectors

    International Nuclear Information System (INIS)

    Marques, Andre Luis Ferreira

    1995-01-01

    This dissertation has aimed at developing a neutron flux measurement technique by means of detectors activation analysis. The main task of this work was the implementation of this thermal neutron flux measurement technique, using gold wires as activation detectors in the IPEN/MB-01 reactor core. The neutron thermal flux spatial distribution was obtained by gold wire activation technique, with wire diameters of 0.125 mm and 0.250 mm in seven selected reactor experimental channels. The values of thermal flux were about 10 9 neutrons/cm 2 .s. This experiment has been the first one conducted with gold wires in the IPEN/MB-01 reactor, being this technique implemented for use by experiments in flux mapping of the core

  20. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  1. Procedure for operating a heavy water cooled power reactor

    International Nuclear Information System (INIS)

    Rau, P.; Kumpf, H.

    1981-01-01

    Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

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

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

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

  3. Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

    Science.gov (United States)

    Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

    2015-01-01

    Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

  4. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  5. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    International Nuclear Information System (INIS)

    Bromley, B.P.; Edwards, G.W.R.; Sambavalingam, P.

    2016-01-01

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  6. Development of supercritical water reactors in Russia and abroad

    International Nuclear Information System (INIS)

    Glebov, A.P.; Klushin, A.V.

    2014-01-01

    The results of Russian and foreign studies on the water-cooled high critical parameters reactors are analyzed. Developments on this subject are conducted in more than 15 countries. The advantages of WWER- SCP and characteristics of experimental reactor of WWER-SCP-30 are discussed. It is noted that priority task is to develop a reactor with thermal neutron spectrum with a subsequent transition to the reactor with a fast neutron spectrum [ru

  7. Analysis of JKT01 Neutron Flux Detector Measurements In RSG-GAS Reactor Using LabVIEW

    Science.gov (United States)

    Rokhmadi; Nur Rachman, Agus; Sujarwono; Taryo, Taswanda; Sunaryo, Geni Rina

    2018-02-01

    The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It is therefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEW was developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEW available in the PC. By using this simulation program, it is successful to perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation results showed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, 1.5×,1.7× and 2.0×.

  8. The Water Suitcase of Migrants: Assessing Virtual Water Fluxes Associated to Human Migration.

    Science.gov (United States)

    Metulini, Rodolfo; Tamea, Stefania; Laio, Francesco; Riccaboni, Massimo

    2016-01-01

    Disentangling the relations between human migrations and water resources is relevant for food security and trade policy in water-scarce countries. It is commonly believed that human migrations are beneficial to the water endowments of origin countries for reducing the pressure on local resources. We show here that such belief is over-simplistic. We reframe the problem by considering the international food trade and the corresponding virtual water fluxes, which quantify the water used for the production of traded agricultural commodities. By means of robust analytical tools, we show that migrants strengthen the commercial links between countries, triggering trade fluxes caused by food consumption habits persisting after migration. Thus migrants significantly increase the virtual water fluxes and the use of water in the countries of origin. The flux ascribable to each migrant, i.e. the "water suitcase", is found to have increased from 321 m3/y in 1990 to 1367 m3/y in 2010. A comparison with the water footprint of individuals shows that where the water suitcase exceeds the water footprint of inhabitants, migrations turn out to be detrimental to the water endowments of origin countries, challenging the common perception that migrations tend to relieve the pressure on the local (water) resources of origin countries.

  9. Nuclear fuel for light water reactors

    International Nuclear Information System (INIS)

    Etemad, A.

    1976-01-01

    The goal of the present speech is to point out some of the now-a-day existing problems related to the fuel cycle of light water reactors and to foresee their present and future solutions. Economical aspects of nuclear power generation have been considerably improving, partly through technological advancements and partly due to the enlargement of unit capacity. The fuel cycle, defined in the course of this talk, discusses the exploration, mining, ore concentration, purification, conversion, enrichment, manufacturing of fuel elements, their utilization in a reactor, their discharge and subsequent storage, reprocessing, and their re-use or disposal. Uranium market in the world and the general policy of several uranium owning countries are described. The western world requirement for uranium until the year 2000, uranium resources and the nuclear power programs in the United States, Australia, Canada, South Africa, France, India, Spain, and Argentina are discussed. The participation of Iran in a large uranium enrichment plant based on French diffusion technology is mentioned

  10. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2005-01-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

  11. Hydrogen behavior in light-water reactors

    International Nuclear Information System (INIS)

    Berman, M.; Cummings, J.C.

    1984-01-01

    The Three Mile Island accident resulted in the generation of an estimated 150 to 600 kg of hydrogen, some of which burned inside the containment building, causing a transient pressure rise of roughly 200 kPa (2 atm). With this accident as the immediate impetus and the improved safety of reactors as the long-term goal, the nuclear industry and the Nuclear Regulatory Commission initiated research programs to study hydrogen behavior and control during accidents at nuclear plants. Several fundamental questions and issues arise when the hydrogen problem for light-water-reactor plants is examined. These relate to four aspects of the problem: hydrogen production; hydrogen transport, release, and mixing; hydrogen combustion; and prevention or mitigation of hydrogen combustion. Although much has been accomplished, some unknowns and uncertainties still remain, for example, the rate of hydrogen production during a degraded-core or molten-core accident, the rate of hydrogen mixing, the effect of geometrical structures and scale on combustion, flame speeds, combustion completeness, and mitigation-scheme effectiveness. This article discusses the nature and extent of the hydrogen problem, the progress that has been made, and the important unresolved questions

  12. Theoretical analysis of nuclear reactors (Phase I), I-V, Part V, Determining the fine flux distribution

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1962-07-01

    Mono energetic neutron transport equation was solved by Carlson numerical method in cylindrical geometry. S n code was developed for the digital computer ZUSE Z23. Neutron flux distribution was determined for the RA reactor cell by applying S 4 approximation. Reactor cell was treated as D 2 O-U-D 2 O system. Time of iteration was 185 s [sr

  13. Feedwater processing method in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izumitani, M; Tanno, K

    1976-09-06

    The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

  14. Automated procedure for selection of optimal refueling policies for light water reactors

    International Nuclear Information System (INIS)

    Lin, B.I.; Zolotar, B.; Weisman, J.

    1979-01-01

    An automated procedure determining a minimum cost refueling policy has been developed for light water reactors. The procedure is an extension of the equilibrium core approach previously devised for pressurized water reactors (PWRs). Use of 1 1/2-group theory has improved the accuracy of the nuclear model and eliminated tedious fitting of albedos. A simple heuristic algorithm for locating a good starting policy has materially reduced PWR computing time. Inclusion of void effects and use of the Haling principle for axial flux calculations extended the nuclear model to boiling water reactors (BWRs). A good initial estimate of the refueling policy is obtained by recognizing that a nearly uniform distribution of reactivity provides low-power peaking. The initial estimate is improved upon by interchanging groups of four assemblies and is subsequently refined by interchanging individual assemblies. The method yields very favorable results, is simpler than previously proposed BWR fuel optimization schemes, and retains power cost as the objective function

  15. Improvements to measuring water flux in the vadose zone.

    Science.gov (United States)

    Masarik, Kevin C; Norman, John M; Brye, Kristofor R; Baker, John M

    2004-01-01

    Evaluating the impact of land use practices on ground water quality has been difficult because few techniques are capable of monitoring the quality and quantity of soil water flow below the root zone without disturbing the soil profile and affecting natural flow processes. A recently introduced method, known as equilibrium tension lysimetry, was a major improvement but it was not a true equilibrium since it still required manual intervention to maintain proper lysimeter suction. We addressed this issue by developing an automated equilibrium tension lysimeter (AETL) system that continuously matches lysimeter tension to soil-water matric potential of the surrounding soil. The soil-water matric potential of the bulk soil is measured with a heat-dissipation sensor, and a small DC pump is used to apply suction to a lysimeter. The improved automated approach reported here was tested in the field for a 12-mo period. Powered by a small 12-V rechargeable battery, the AETLs were able to continuously match lysimeter suction to soil-water matric potential for 2-wk periods with minimal human attention, along with the added benefit of collecting continuous soil-water matric potential data. We also demonstrated, in the laboratory, methods for continuous measurement of water depth in the AETL, a capability that quantifies drainage on a 10-min interval, making it a true water-flux meter. Equilibrium tension lysimeters have already been demonstrated to be a reliable method of measuring drainage flux, and the further improvements have created a more effective device for studying water drainage and chemical leaching through the soil matrix.

  16. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  17. The heavy water accountancy for research reactors in JAERI

    International Nuclear Information System (INIS)

    Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

    1998-11-01

    The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

  18. Temporal variation of the neutron flux in the carousel facility of a TRIGA reactor

    International Nuclear Information System (INIS)

    Jacimovic, R.; Stegnar, P.; Trkov, A.

    2003-01-01

    In this work we focused on identifying quantitatively the temporal (time-dependent) variation of neutron flux in the carousel facility (CF) of TRIGA reactor at the 'Jozef Stefan' Institute (IJS) for core No. 176, set up in April 2002. The measurements are based on neutron detectors (ionisation chambers), which surround the graphite reflector of the reactor core. In principle, the variations of the neutron flux produce a systematic error in the results obtained by absolute or 'quasi' absolute measuring techniques (such as neutron activation analysis (NAA) by the k 0 -standardization method), which assume constant conditions during irradiation. The results of our study show that for typical irradiation of 20 hours in channels of the CF aligned in the direction of the ionisation chamber (safety channel) the time-dependent variation of the neutron flux is about 6-8%. In the k 0 method, which we are using for routine work at the IJS, this variation introduced a systematic error in the results up to 4.6%, depending on the half-life of investigated radionuclide. (author)

  19. Improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell

    International Nuclear Information System (INIS)

    Bosevski, T.

    1986-01-01

    An improved collision probability method for thermal-neutron-flux calculation in a cylindrical reactor cell has been developed. Expanding the neutron flux and source into a series of even powers of the radius, one' gets a convenient method for integration of the one-energy group integral transport equation. It is shown that it is possible to perform an analytical integration in the x-y plane in one variable and to use the effective Gaussian integration over another one. Choosing a convenient distribution of space points in fuel and moderator the transport matrix calculation and cell reaction rate integration were condensed. On the basis of the proposed method, the computer program DISKRET for the ZUSE-Z 23 K computer has been written. The suitability of the proposed method for the calculation of the thermal-neutron-flux distribution in a reactor cell can be seen from the test results obtained. Compared with the other collision probability methods, the proposed treatment excels with a mathematical simplicity and a faster convergence. (author)

  20. Technological readiness of evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Juhn, P.E.

    1999-01-01

    Nuclear energy has evolved to a mature industry that supplies over 16% of the world's electricity, and it represents an important option for meeting the global energy demands of the coming century in an environmentally acceptable manner. New, evolutionary water cooled reactor designs that build on successful performance of predecessors have been developed; these designs have generally been guided by wishes to reduce cost, to improve availability and reliability, and to meet increasingly stringent safety objectives. These three aspects are important factors in what has been called technological readiness for an expanded deployment of nuclear power; a major increase in utilization of nuclear power will only occur if it is economically competitive, and meets safety expectations. To this end, the industry will also have to maintain or improve the public perception of nuclear power as a benign, economical and reliable energy source. (author)

  1. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

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

  2. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  3. Light water reactor lower head failure analysis

    International Nuclear Information System (INIS)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

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

  4. Reprocessing technology for present water reactor fuels

    International Nuclear Information System (INIS)

    McMurray, P.R.

    1977-01-01

    The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

  5. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

  6. Corrosion problems in light water nuclear reactors

    International Nuclear Information System (INIS)

    Berry, W.E.

    1984-01-01

    The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers

  7. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-10-01

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

  8. A review of the UKAEA interest in heavy water reactors

    International Nuclear Information System (INIS)

    Symes, R.J.

    1983-01-01

    The chapter commences with a brief account of the history of heavy water production and then begins the story of the British use of this moderator in power reactors. This is equated with the introduction and development of the tube reactor as a distinct and important form of reactor construction in contrast with the perhaps better known vessel design that has tended to dominate reactor engineering to date. The account thus includes a succession of reactor designs including the gas and steam cooled heavy water systems in addition to the steam-generating heavy water reactor. The SGHWR was demonstrated by the construction of a substantial prototype, which continues in operation as a flexible and reliable electricity-generating plant. It was also, for a time, identified as the system to be used for Britain's third reactor programme. Today the successful Canadian CANDU power reactors represent the only penetration of heavy water reactor technology into large scale electricity generation. The range of research and experimental reactors using heavy water in their cores is reviewed. (author)

  9. Water flux in animals: analysis of potential errors in the tritiated water method

    Energy Technology Data Exchange (ETDEWEB)

    Nagy, K.A.; Costa, D.

    1979-03-01

    Laboratory studies indicate that tritiated water measurements of water flux are accurate to within -7 to +4% in mammals, but errors are larger in some reptiles. However, under conditions that can occur in field studies, errors may be much greater. Influx of environmental water vapor via lungs and skin can cause errors exceeding +-50% in some circumstances. If water flux rates in an animal vary through time, errors approach +-15% in extreme situations, but are near +-3% in more typical circumstances. Errors due to fractional evaporation of tritiated water may approach -9%. This error probably varies between species. Use of an inappropriate equation for calculating water flux from isotope data can cause errors exceeding +-100%. The following sources of error are either negligible or avoidable: use of isotope dilution space as a measure of body water volume, loss of nonaqueous tritium bound to excreta, binding of tritium with nonaqueous substances in the body, radiation toxicity effects, and small analytical errors in isotope measurements. Water flux rates measured with tritiated water should be within +-10% of actual flux rates in most situations.

  10. Water flux in animals: analysis of potential errors in the tritiated water method

    International Nuclear Information System (INIS)

    Nagy, K.A.; Costa, D.

    1979-03-01

    Laboratory studies indicate that tritiated water measurements of water flux are accurate to within -7 to +4% in mammals, but errors are larger in some reptiles. However, under conditions that can occur in field studies, errors may be much greater. Influx of environmental water vapor via lungs and skin can cause errors exceeding +-50% in some circumstances. If water flux rates in an animal vary through time, errors approach +-15% in extreme situations, but are near +-3% in more typical circumstances. Errors due to fractional evaporation of tritiated water may approach -9%. This error probably varies between species. Use of an inappropriate equation for calculating water flux from isotope data can cause errors exceeding +-100%. The following sources of error are either negligible or avoidable: use of isotope dilution space as a measure of body water volume, loss of nonaqueous tritium bound to excreta, binding of tritium with nonaqueous substances in the body, radiation toxicity effects, and small analytical errors in isotope measurements. Water flux rates measured with tritiated water should be within +-10% of actual flux rates in most situations

  11. Residual fluxes of water, salt and suspended sediment in the Beypore Estuary

    Digital Repository Service at National Institute of Oceanography (India)

    AnilKumar, N.; Revichandran, C.; Sankaranarayanan, V.N.; Josanto, V.

    The monthly trends of the residual fluxes of salt and water and the transportation of suspended sediments in the Beypore estuarine system, Kerala, India were examined. At the river mouth the water flux was directed seaward during the postmonsoon...

  12. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Bolme, A.B.

    1996-01-01

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  13. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Meneley, D.A.; Olmstead, R.A.; Yu, A.M.; Dastur, A.R.; Yu, S.K.W.

    1994-01-01

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  14. Computer simulation of the NASA water vapor electrolysis reactor

    Science.gov (United States)

    Bloom, A. M.

    1974-01-01

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

  15. On the slimeless water operation in the RBMK type reactors

    International Nuclear Information System (INIS)

    Margulova, T.Kh.; Mamet, V.A.; Nikitina, I.S.; Karakhanyan, L.N.

    1988-01-01

    Water chemistry conditions of the operating RBMK-1000 and RBMK-1500 units are analysed. Inevitability of iron oxide deposits in RBMK-1000 and particularly in RBMK-1500 reactors is demonstrated. Organization of a new slimeless correcting water chemistry for RBMK-1000 and RBMK-1500 reactors is recommended

  16. Development of light water reactors and subjects for hereafter

    International Nuclear Information System (INIS)

    Murao, Yoshio

    1995-01-01

    As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)

  17. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    Queiser, H.

    1976-01-01

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-10

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

  19. Suppression device for the reactor water level lowering

    International Nuclear Information System (INIS)

    Kasuga, Hajime; Kasuga, Hiroshi.

    1984-01-01

    Purpose: To suppress the lowering in the reactor water level so as to avoid unnecessary actuation of ECCS upon generation of transient changes which forecasts the lowering of the reactor water level in a BWR type reactor. Constitution: There are provided a water level suppression signal generator for generating a water level suppression signal upon generation of a transient change signal which forecasts the water level lowering in a nuclear reactor and a recycling flow rate controller that applies a recycling flow rate control signal to a recycling pump drive motor by the water level lowering suppression signal. The velocity of the recycling pump is controlled by a reactor scram signal by way of the water level lowering suppresion signal generator and a recycling flow rate controller. Then, the recycling reactor core flow rate is decreased and the void amount in the reactor is transiently increased where the water level tends to increase. Accordingly, the water level lowering by the scram is moderated by the increasing tendency of the water level. (Ikeda, J.)

  20. Measurement and calculation of spatial and energetic neutron flux in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    Bittelli, U.D.

    1988-01-01

    This work presents spatial and energetic flux distribution measured in the IEA-R1 reactor core. The thermal neutron flux was measured by gold activation foils (bare and covered with cadmium) in the fuel element number 108 (reaction: 197 Au(n,γ) 198 Au) at 451W overall reactor power. The fast neutron flux was measured by indium activation foils (reaction: 115 In(n,n') 115m In) in the fuel elements number 94 at 4510W overall reactor power. The neutron energy spectrum was adjusted by SAND II code with the data produced by the irradiation of seven activation detectors in the fuel element number 94 at 4510 W overall reactor power. The following reactions were used: 58 Fe(n,γ) 59 Fe, 232 Th(n,γ) 233 Th, 197 Au(n,γ) 198 Au, 59 Co(n,γ) 60 Co, 54 Fe(n,p) 54 Mn, 24 Mg(n,p) 24 Na, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc and 115 In(n,n') 115m In. The experimental results compared to those obtained by CITATION (spatial distribution flux) and HAMMER (energetic distribution flux) code, showed good agreement. The results presented in this work are a good contribution for a better knowledge of spatial and energetic neutron flux distribution in the IEA-R1 reactor core, besides that the experimental procedure is easily applicable to another situations. (autor) [pt