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Sample records for neutron-irradiated pressure vessel

  1. Positron annihilation studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Brauer, G.; Liszkay, L.; Molnar, B.

    1988-01-01

    Several annealing studies by positron annihilation (Doppler broadening, lifetime) on neutron irradiated Cr-Mo-V reactor pressure vessel steels (Soviet type 15Kh2MFA) regarding the influences of irradiation temperature, fluence of fast neutrons as well as different impurity contents are presented and discussed. A possibility of explaining the positron annihilation data by irradiation induced carbide formation is proposed. (author)

  2. Neutron irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Steele, L.E.

    1975-01-01

    The reliability of nuclear power plants depends on the proper functioning of complex components over the whole life on the plant. Particular concern for reliability is directed to the primary pressure boundary. This report focuses on the portion of the primary system exposed to and significantly affected by neutron radiation. Experimental evidence from research programmes and from reactor surveillance programmes has indicated radiation embrittlement of a magnitude sufficient to raise doubts about reactor pressure vessel integrity. The crucial nature of the primary vessel function heightens the need to be alert to this problem, to which, fortunately, there are positive aspects: for example, steels have been developed which are relatively immune to radiation embrittlement. Further, awareness of such embrittlement has led to designs which can accomodate this factor. The nature of nuclear reactors, of the steels used in their construction, and of the procedures for interpreting embrittlement and minimizing the effects are reviewed with reference to the reactors that are expected to play a major role in electric power production from now to about the turn of the century. The report is intended as a manual or guidebook; the aim has been to make each chapter or major sub-division sufficiently comprehensive and self-contained for it to be understood and read independently of the rest of the book. At the same time, it is hoped that the whole is unified enough to make a complete reading useful and interesting to the several classes of reader that are involved with only specific aspects of the topic

  3. Characterization of phosphorus segregation in neutron-irradiated Russian pressure vessel steel weld

    International Nuclear Information System (INIS)

    Miller, M.K.; Jayaram, R.; Russell, K.F.

    1995-01-01

    An atom probe field ion microscopy characterization of three Russian pressure vessel steels has been performed. Field ion micrographs of several lath boundaries have indicated that they are decorated with a semicontinuous film of discrete brightly-imaging precipitates that were identified as molybdenum carbonitrides. In addition, extremely high phosphorus levels were measured at the lath boundaries. The phosphorus was found to be confined to an extremely narrow region indicative of monolayer type segregation. The phosphorus coverage determined from the atom probe results of the unirradiated materials agree with predictions based on McLean's equilibrium model of grain boundary segregation. The boundary phosphorus coverage of a neutron-irradiated weld material was significantly higher than in the unirradiated material. Ultrafine darkly-imaging copper- and phosphorus-enriched precipitates were also observed in the matrix of the neutron-irradiated material. (orig.)

  4. Neutron irradiation effects in reactor pressure vessel steels and weldments. Working document

    International Nuclear Information System (INIS)

    1998-10-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. A separate abstract was prepared for the introduction and for each of the eleven chapters, which are: 1. Reactor Pressure Vessel Design, 2. Reactor Pressure Materials, 3. WWER Pressure Vessels, 4. Determination of Mechanical Properties, 5. Neutron Exposure, 6. Methodology of Irradiation Experiments, 7. Effect of Irradiation on Mechanical Properties, 8. Mechanisms of Irradiation Embrittlement, 9. Modelling of Irradiation Damage, 10. Annealing of Irradiation Damage, 11. Safety Assessment using Surveillance Programmes and Data Bases

  5. Cohesive modelling of the fracture of a neutron irradiated pressure vessel steel

    International Nuclear Information System (INIS)

    Gomez, F.J.; Valiente, A.; Elices, M.

    2003-01-01

    The cohesive fracture process zone model was used to account for the neutron irradiation embrittlement of a pressure vessel steel. The tensile testing and fracture of axisymmetrically notched round specimens were numerically modelled assuming a rectangular traction separation law and the irradiation effects were introduced by due modification of this law. The results corroborate those of the experiments performed in a previous work. The cohesive strength and the cohesive energy of the cohesive model were not considered as adjusting parameters, but they were determined from the data of conventional tensile tests and fracture toughness tests on the assumption that the failure of the specimens in these tests also follows the cohesive model

  6. Electron-microscopic investigation of a pressure vessel steel after neutron irradiation

    International Nuclear Information System (INIS)

    Klaar, H.J.

    1975-01-01

    As an introduction, changes in the mechanical properties of pressure vessel steels on neutron irradiation and the causes of radiation embrittlement are discussed. After this, the author describes his own experiments with steel of the composition 0.19% C; 3.88% Ni; 1.57% Cr; 0.51% Mo; 0.2% V. Samples of this material were irradiated in-pile at 300 0 C with various neutron doses. To study the influence of neutron dose, irradiation temperature, and heat treatment on the mechanical properties, tensile tests, notched bar impact bending tests, hardness tests and structural analyses were carried out. The findings are reported. (GSC) [de

  7. Fatigue crack propagation in neutron-irradiated ferritic pressure-vessel steels

    International Nuclear Information System (INIS)

    James, L.A.

    1977-01-01

    The results of a number of experiments dealing with fatigue crack propagation in irradiated reactor pressure-vessel steels are reviewed. The steels included ASTM alloys A302B, A533B, A508-2, and A543, as well as weldments in A543 steel. Fluences and irradiation conditions were generally typical of those experienced by most power reactors. In general, the effect of neutron irradiation on the fatigue crack propagation behavior of these steels was neither significantly beneficial nor significantly detrimental

  8. Effects of neutron irradiation on resistivity of reactor pressure vessel steel

    Science.gov (United States)

    Li, Chengliang; Shu, Guogang; Liu, Yi; Huang, Yili; Chen, Jun; Duan, Yuangang; Liu, Wei

    2018-02-01

    The embrittlement of reactor pressure vessel (RPV) steel owing to fast-neutron irradiation is one of its primary failure mechanisms. In this work, neutron irradiation tests were performed on an RPV steel at a high temperature (565 K) using a neutron irradiation test reactor. In addition, resistivity measurements were performed on the RPV steel both before and after irradiation in a hot laboratory using the four-probe method. The results showed that the resistivity of the RPV steel exhibits nonlinear behaviour with respect to the radiation fluence and that the nonlinearity becomes more pronounced with an increase in the radiation fluence. For instance, when the radiation fluence is 0.1540 dpa and the excitation current is increased from 0.2 mA to 200 mA, the resistivity of the RPV steel decreases by as much as 67.12%. During irradiation embrittlement, the resistivity increases with the fluence. When the radiation fluence is greater than 0.116 dpa, the increase in the resistivity accelerates. When the radiation fluence is less than 0.116 dpa and when an excitation current of 2 mA or 20 mA is used, the relationship between the resistivity and the radiation fluence for the RPV steel is a quadratic one, whereas that between the rate of change in the resistivity and the radiation fluence is a linear one. Thus, the resistivity of RPV steel can be used to characterise its degree of irradiation embrittlement, and resistivity measurements can be employed as a nondestructive evaluation technique for monitoring the degree of irradiation damage experienced by in-service RPV steel.

  9. Nanostructure evolution of neutron-irradiated reactor pressure vessel steels: Revised Object kinetic Monte Carlo model

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Messina, L. [DEN-Service de Recherches de Métallurgie Physique, CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette (France); KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Olsson, P. [KTH Royal Institute of Technology, Roslagstullsbacken 21, SE-114 21 Stockholm (Sweden); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    This work presents a revised set of parameters to be used in an Object kinetic Monte Carlo model to simulate the microstructure evolution under neutron irradiation of reactor pressure vessel steels at the operational temperature of light water reactors (∼300 °C). Within a “grey-alloy” approach, a more physical description than in a previous work is used to translate the effect of Mn and Ni solute atoms on the defect cluster diffusivity reduction. The slowing down of self-interstitial clusters, due to the interaction between solutes and crowdions in Fe is now parameterized using binding energies from the latest DFT calculations and the solute concentration in the matrix from atom-probe experiments. The mobility of vacancy clusters in the presence of Mn and Ni solute atoms was also modified on the basis of recent DFT results, thereby removing some previous approximations. The same set of parameters was seen to predict the correct microstructure evolution for two different types of alloys, under very different irradiation conditions: an Fe-C-MnNi model alloy, neutron irradiated at a relatively high flux, and a high-Mn, high-Ni RPV steel from the Swedish Ringhals reactor surveillance program. In both cases, the predicted self-interstitial loop density matches the experimental solute cluster density, further corroborating the surmise that the MnNi-rich nanofeatures form by solute enrichment of immobilized small interstitial loops, which are invisible to the electron microscope.

  10. Evolution of the microstructure of a French reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Huang, H.F.; Radiguet, B.; Pareige, P. [Groupe de physique des materiaux, UMR CNRS 6634, Universite et INSA de Rouen, avenue de l' Universite, 76801 Saint Etienne du Rouvray (France); Todeschini, P. [EDF, Materials and Mechanics of Component Department, Site des Renardieres-Ecuelles, 77818 Moret-sur-Loing cedex (France); Chas, G. [EDF, Production and Engineering Branch, CEIDRE/DLAB, CNPE de Chinon, BP 23, 37420 Avoine (France)

    2011-07-01

    The microstructure of a low copper French reactor pressure vessel steel, neutron irradiated within the frame of the EDF Surveillance Program of a production reactor, was characterised by atom probe tomography. Specimens were irradiated at low flux (2*10{sup 15} m{sup -2}/s), at 4 different fluences up to 7.6*10{sup 23} m{sup -2}. Atom Probe experiments have revealed the presence of roughly spherical clusters enriched in nickel, manganese, silicon and, in a lesser extent, phosphorus and copper at all irradiation fluences. The chemical composition of these clusters shows no evolution with fluence, as well as their diameter, close to 3 nm. On the contrary, their number density increases linearly with the neutron fluence. Continuous segregation of the elements found in the clusters is also observed along dislocation lines, with similar enrichments. (authors)

  11. 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

  12. Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Oarai Research and Development Center, Japan Atomic Energy Agency, Oarai, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Toyama, T.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Yamaguchi, Y.; Onizawa, K. [Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

    2014-06-01

    The microstructures and the hardness of stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation at a dose of 7.2 × 10{sup 19} n cm{sup −2} (E > 1 MeV) and a flux of 1.1 × 10{sup 13} n cm{sup −2} s{sup −1} at 290 °C were investigated by atom probe tomography and by a nanoindentation technique. To isolate the effects of the neutron irradiation, we compared the results of the measurements of the neutron-irradiated samples with those from a sample aged at 300 °C for a duration equivalent to that of the irradiation. The Cr concentration fluctuation was enhanced in the δ-ferrite phase of the irradiated sample. In addition, enhancement of the concentration fluctuation of Si, which was not observed in the aged sample, was observed. The hardening in the δ-ferrite phase occurred due to both irradiation and aging; however, the hardening of the irradiated sample was more than that expected from the Cr concentration fluctuation, which suggested that the Si concentration fluctuation and irradiation-induced defects were possible origins of the additional hardening.

  13. Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, T., E-mail: takeuchi.tomoaki@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashiibaraki-gun, Ibaraki 311-1393 (Japan); Kakubo, Y.; Matsukawa, Y.; Nozawa, Y.; Nagai, Y. [The Oarai Center, Institute for Materials Research, Tohoku University, Oarai, Ibaraki 311-1313 (Japan); Nishiyama, Y.; Katsuyama, J.; Onizawa, K. [Japan Atomic Energy Agency, 2-4 Shirakata-Shirane, Tokai, Naka-gun, Ibaraki 319-1195 (Japan); Suzuki, M. [Japan Atomic Energy Agency, 4002 Narita, Oarai, Higashiibaraki-gun, Ibaraki 311-1393 (Japan)

    2013-11-15

    Microstructural changes in the stainless steel weld overlay cladding of reactor pressure vessels subjected to neutron irradiation with a fluence of 7.2 × 10{sup 23} n m{sup −2} (E > 1 MeV) and a flux of 1.1 × 10{sup 17} n m{sup −2} s{sup −1} at 290 °C were investigated by atom probe tomography. The results showed a difference in the microstructural changes that result from neutron irradiation and thermal aging. Neutron irradiation resulted in the slight progression of Cr spinodal decomposition and an increase in the fluctuation of the Si, Ni, and Mn concentrations in the ferrite phases, with formation of γ′-like clusters in the austenite phases. On the other hand, thermal aging resulted in the considerable progression of the Cr spinodal decomposition, formation of G-phases, and a decrease in the Si and an increase in the Ni and Mn concentration fluctuations at the matrix in the ferrite phases, without the microstructural changes in the austenite phases.

  14. Effect of neutron irradiation on the mechanical properties of weld overlay cladding for reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Tobita, Tohru, E-mail: tobita.tohru@jaea.go.jp; Udagawa, Makoto; Chimi, Yasuhiro; Nishiyama, Yutaka; Onizawa, Kunio

    2014-09-15

    This study investigates the effects of high fluence neutron irradiation on the mechanical properties of two types of cladding materials fabricated using the submerged-arc welding and electroslag welding methods. The tensile tests, Charpy impact tests, and fracture toughness tests were conducted before and after the neutron irradiation with a fluence of 1 × 10{sup 24} n/m{sup 2} at 290 °C. With neutron irradiation, we could observe an increase in the yield strength and ultimate strength, and a decrease in the total elongation. All cladding materials exhibited ductile-to-brittle transition behavior during the Charpy impact tests. A reduction in the Charpy upper-shelf energy and an increase in the ductile-to-brittle transition temperature was observed with neutron irradiation. There was no obvious decrease in the elastic–plastic fracture toughness (J{sub Ic}) of the cladding materials upon irradiation with high neutron fluence. The tearing modulus was found to decrease with neutron irradiation; the submerged-arc-welded cladding materials exhibited low J{sub Ic} values at high temperatures.

  15. Irradiation embrittlement of reactor pressure vessel steels (Report on analysis of the behaviour of advanced reactor pressure vessel steel under neutron irradiation)

    International Nuclear Information System (INIS)

    Sivaramakrishnan, K.S.; Chatterjee, S.; Anantharaman, S.; Balakrishnan, K.S.; Viswanathan, U.K.

    1984-01-01

    The International Working Group on Reliability of Reactor Pressure Components (IWG-RRPC) sponsored by IAEA launched a Coordinated Research Programme in 1977 to investigate the behaviour of Advanced Pressure Vessel Steels under neutron irradiation. IWG-RRPC included provision of various materials produced by organisation in Federal Republic of Germany, France, Japan and a reference Steel from United States of America to those countries participating in the programme. India joined the International Programme in 1977. The scope of study was mainly to evaluate the neutron irradiation induced change in nil ductility transition temperature and tensile properties on some selected materials. The required quantity of these materials was received from Japan, France, Federal Republic of Germany and United States of America. The paper reports the results obtained from the investigations carried out in the Radiometallurgy Division of Bhabha Atomic Research Centre, Trombay, Bombay, India, on Japanese plate, Japanese forging, French plate, French forging and French weld. Due to other important commitments of the CIRUS reactor, in which the irradiation was carried out, it was not possible to complete the entire programme at this time. (author)

  16. Correlating radiation exposure with embrittlement: Comparative studies of electron- and neutron-irradiated pressure vessel alloys

    International Nuclear Information System (INIS)

    Alexander, D. E.; Rehn, L. E.; Odette, G. R.; Lucas, G. E.; Klingensmith, D.; Gragg, D.

    1999-01-01

    Comparative experiments using high energy (10 MeV) electrons and test reactor neutrons have been undertaken to understand the role that primary damage state has on hardening (embrittlement) induced by irradiation at 300 C. Electrons produce displacement damage primarily by low energy atomic recoils, while fast neutrons produce displacements from considerably higher energy recoils. Comparison of changes resulting from neutron irradiation, in which nascent point defect clusters can form in dense cascades, with electron irradiation, where cascade formation is minimized, can provide insight into the role that the in-cascade point defect clusters have on the mechanisms of embrittlement. Tensile property changes induced by 10 MeV electrons or test reactor neutron irradiations of unalloyed iron and an Fe-O.9 wt.% Cu-1.0 wt.% Mn alloy were examined in the damage range of 9.0 x 10 -5 dpa to 1.5 x 10 -2 dpa. The results show the ternary alloy experienced substantially greater embrittlement in both the electron and neutron irradiate samples relative to unalloyed iron. Despite their disparate nature of defect production similar embrittlement trends with increasing radiation damage were observed for electrons and neutrons in both the ternary and unalloyed iron

  17. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels(I) (1st progress report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jun Hwa; Lee, Bong Sang; Park, Duck Gun; Byun, Tak Sang; Kim, Joo Hag; Oh, Yong Jun; Yoon, Ji Hyun; Chi, Sei Hwan; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    The SA508-3 reactor pressure vessel materials degrade due to the application at high temperature, high pressure, and neutron irradiation. In the present study it is planned to examine the effects of neutron irradiation on the properties for assessing the integrity of domestic reactors. The key tests are the Charpy impact test, tensile test, static and dynamic fracture toughness test, J-R test. The additional tests for obtaining basic material properties, such as micro-hardness, microstructural properties, small punch energy etc., are also performed. The irradiation tests are being performed at HANARO of KAERI through the instrumented capsules designed by KAERI and the post-irradiation tests are being performed at IMEF(Irradiated Material Evaluation Facility) of material (UCN-4), Si+Al (YGN-5), UCN-4 weld metal, and UCN-4 HAZ. In the irradiation test the temperature should be controlled in the range of 290 {+-} 10 deg C and the test materials would be irradiated to 2 to 3 neutron fluence levels including the end-of-life fluence. The status of performing this project is that (1) the key data on mechanical properties, mainly related to the fracture toughness, of the unirradiated materials have been obtained, (2) the irradiation of the 1st instrumented capsule, a preliminary test capsule containing miniature specimens, has been completed and is being stored for testing in IMEF, and (3) the 2nd instrumented capsule is being manufactured and will be irradiated in the beginning or 1999. This report includes mainly the experimental methods and results. The status of the design and manufacturing of the instrumented capsules and specimens was also briefly described. (author). 13 refs., 15 figs., 10 tabs.

  18. Structural characterization of nanoscale intermetallic precipitates in highly neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Sprouster, D.J.; Sinsheimer, J.; Dooryhee, E.; Ghose, S.K.; Wells, P.; Stan, T.; Almirall, N.; Odette, G.R.; Ecker, L.E.

    2016-01-01

    Massive, thick-walled pressure vessels are permanent nuclear reactor structures that are exposed to a damaging flux of neutrons from the adjacent core. The neutrons cause embrittlement of the vessel steel that grows with dose (fluence), as manifested by an increasing ductile-to-brittle fracture transition temperature. Extending reactor life requires demonstrating that large safety margins against brittle fracture are maintained at the higher neutron fluence associated with beyond 60 years of service. Here synchrotron-based x-ray diffraction and small angle x-ray scattering measurements are used to characterize highly embrittling nm-scale Mn–Ni–Si precipitates that develop in the irradiated steels at high fluence. These precipitates lead to severe embrittlement that is not accounted for in current regulatory models. Application of the complementary techniques has, for the very first time, successfully identified the crystal structures of the nanoprecipitates, while also yielding self-consistent compositions, volume fractions and size distributions.

  19. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR

    International Nuclear Information System (INIS)

    Moranchel y R, M.

    2012-01-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  20. Atom Probe Tomography Characterization of the Solute Distributions in a Neutron-Irradiated and Annealed Pressure Vessel Steel Weld

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.K.

    2001-01-30

    A combined atom probe tomography and atom probe field ion microscopy study has been performed on a submerged arc weld irradiated to high fluence in the Heavy-Section Steel irradiation (HSSI) fifth irradiation series (Weld 73W). The composition of this weld is Fe - 0.27 at. % Cu, 1.58% Mn, 0.57% Ni, 0.34% MO, 0.27% Cr, 0.58% Si, 0.003% V, 0.45% C, 0.009% P, and 0.009% S. The material was examined after five conditions: after a typical stress relief treatment of 40 h at 607 C, after neutron irradiation to a fluence of 2 x 10{sup 23} n m{sup {minus}2} (E > 1 MeV), and after irradiation and isothermal anneals of 0.5, 1, and 168 h at 454 C. This report describes the matrix composition and the size, composition, and number density of the ultrafine copper-enriched precipitates that formed under neutron irradiation and the change in these parameters with post-irradiation annealing treatments.

  1. A new method for improving the reliability of fracture toughness surveillance of nuclear pressure vessel by neutron irradiated embrittlement

    International Nuclear Information System (INIS)

    Zhang Xinping; Shi Yaowu

    1992-01-01

    In order to obtain more information from neutron irradiated sample specimens and raise the reliability of fracture toughness surveillance test, it has more important significance to repeatedly exploit the broken Charpy-size specimen which had been tested in surveillance test. In this work, on the renewing design and utilization for Charpy-size specimens, 9 data of fracture toughness can be gained from one pre-cracked side-grooved Charpy-size specimen while at the preset usually only 1 to 3 data of fracture toughness can be obtained from one Chharpy-size specimen. Thus, it is found that the new method would obviously improve the reliability of fracture toughness surveillance test and evaluation. Some factors which affect the reasonable design of pre-cracked deep side-groove Charpy-size compound specimen have been discussed

  2. Correlation of mechanical property changes in neutron-irradiated pressure vessel steels on the basis of spectral effects

    International Nuclear Information System (INIS)

    Heinisch, H.L.

    1991-01-01

    Comparisons are made of tensile data on specimens of A212B and A302B pressure vessel steels irradiated at low temperatures (40-90degC) and to low doses (<0.1 dpa) with 14 MeV D-T fusion neutrons in the Rotating Target Neutron Source (RTNS-II), with fission reactor neutrons in the Omega West Reactor (OWR) and the Oak Ridge Research Reactor (ORR), and with the highly thermal spectrum at the pressure vessel surveillance positions of the High Flux Isotope Reactor (HFIR). For each neutron spectrum, damage cross sections are determined for several defect production functions derived from atomistic computer simulations of collision cascades. Displacements per atom (dpa) and the numbers of freely migrating defects are tested as damage correlation parameters for the tensile data. The data from RTNS-II, OWR and ORR correlate fairly well when compared on the basis of dpa, but the data from HFIR show only about one sixth as many dpa are needed to produce the same radiation-induced yield stress changes as in the other neutron spectra. In the HFIR surveillance position a significant fraction of the displacements is produced by recoils resulting from thermal neutron captures. Having energies of about 400 eV, these recoils are much more efficient per unit energy at producing freely migrating defects than the high energy recoils responsible for most of the displacements in the other neutron spectra considered. A significantly better correlation of data from HFIR with those from the other spectra is achieved when the property changes are compared on the basis of the production of freely migrating self-interstitial defects. (orig./MM)

  3. The changes of the structural, magnetic, and mechanical properties in a reactor pressure vessel steel neutron-irradiated at 70 .deg. C

    CERN Document Server

    Park, D G; Jang, K S; Jung, M M; Kim, G M

    1999-01-01

    The irradiation embrittlement of reactor-pressure-vessel steel has been one of the main safety concerns in nuclear power plants. In the present study, an SA508-3 RPV steel was irradiated by neutrons with various fluences up to 10 sup 1 sup 8 n/cm sup 2 (E>=1MeV) at a temperature of approximately 70 .deg. C. The irradiation responses of the structural, the magnetic, and the mechanical properties of the steel were investigated by means of X-ray diffraction, Moessbauer spectroscopy, magnetic Barkhausen noise, and micro-Vickers hardness measurements. The transitions of all of these parameters occurred above a neutron does of 10 sup 1 sup 6 n/cm sup 2. The results of the X-ray and the Moessbauer experiments revealed that neutron irradiation led to the possibility of partial amorphization in the investigated RPV steel. The changes of the physical and the mechanical properties were discussed in terms of irradiation-induced cascade damage of crystalline materials.

  4. The Change of the Seebeck Coefficient Due to Neutron Irradiation and Thermal Fatigue of Nuclear Reactor Pressure Vessel Steel and its Application to the Monitoring of Material Degradation

    International Nuclear Information System (INIS)

    Niffenegger, M.; Reichlin, K.; Kalkhof, D.

    2002-05-01

    The monitoring of material degradation, that might be caused by neutron irradiation and thermal fatigue, is an important topic in lifetime extension of nuclear power plants. We therefore investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel steel. The Seebeck coefficient (SC) of several irradiated Charpy specimens made from Japanese JRQ-steel were measured. The specimens suffered a fluence from 0 up to 4.5 x 10 19 neutrons per cm 2 with energies higher than 1 MeV. The measured changes of the SC within this range were about 500 nV, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.55 x 10 19 neutrons per cm 2 . We obtained a linear dependency between the SC and the temperature shift ΔT 41 of the Charpy-Energy- Temperature curve which is widely used to characterize material embrittlement. Similar measurements were performed on specimens made from the widely used austenitic steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying a cyclic strain amplitude of 0.28%. For this kind of fatigue the observed change of SC was somewhat smaller than for the irradiated specimens. Further investigations were made to quantify the size of the gage volume in which the thermoelectric power is generated. It appeared that the information gathered from a Thermo Electric Power (TEP) measurement is very local. To overcome this problem we propose a novel TEP-method using a Thermoelectric Scanning Microscope (TSM). We finally conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. A part of this disadvantage can

  5. An object kinetic Monte Carlo model for the microstructure evolution of neutron-irradiated reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Messina, Luca; Olsson, Paer [KTH Royal Institute of Technology, Stockholm (Sweden); Chiapetto, Monica [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium); Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Becquart, Charlotte S. [Unite Materiaux et Transformations (UMET), UMR 8207, Universite de Lille 1, ENSCL, Villeneuve d' Ascq (France); Malerba, Lorenzo [SCK - CEN, Nuclear Materials Science Institute, Mol (Belgium)

    2016-11-15

    This work presents a full object kinetic Monte Carlo framework for the simulation of the microstructure evolution of reactor pressure vessel (RPV) steels. The model pursues a ''gray-alloy'' approach, where the effect of solute atoms is seen exclusively as a reduction of the mobility of defect clusters. The same set of parameters yields a satisfactory evolution for two different types of alloys, in very different irradiation conditions: an Fe-C-MnNi model alloy (high flux) and a high-Mn, high-Ni RPV steel (low flux). A satisfactory match with the experimental characterizations is obtained only if assuming a substantial immobilization of vacancy clusters due to solute atoms, which is here verified by means of independent atomistic kinetic Monte Carlo simulations. The microstructure evolution of the two alloys is strongly affected by the dose rate; a predominance of single defects and small defect clusters is observed at low dose rates, whereas larger defect clusters appear at high dose rates. In both cases, the predicted density of interstitial loops matches the experimental solute-cluster density, suggesting that the MnNi-rich nanofeatures might form as a consequence of solute enrichment on immobilized small interstitial loops, which are invisible to the electron microscope. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  6. Reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Suzuki, K.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. Chapter 3 offers a detailed treatment of the selection criteria and properties of reactor pressure vessel materials. The main attention is directed towards steel and ingot making and the subsequent material processing

  7. Analysis of the microstructural evolution of the damage by neutron irradiation in the pressure vessel of a nuclear power reactor BWR; Analisis de la evolucion microestructural del dano por irradiacion neutronica en la vasija de presion de un reactor nuclear de potencia BWR

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y R, M.

    2012-07-01

    Nuclear reactor pressure vessel type BWR, installed in Mexico and in many other countries, are made of an alloy of low carbon steel. The American Society for Testing and Materials (Astm) classifies this alloy as A533-B, class 1. Both the vessel and other internal structures are continuously exposed to the neutron flux from the reactions of fission in nuclear fuel. A large number of neutrons reach the vessel and penetrate certain depth depending on their energy. Its penetration in the neutron collides with the nuclei of the atoms out of their positions in the crystal lattice of steel, producing vacancies, interstitial, segregations, among other defects, capable of affecting its mechanical properties. Analyze the micro-structural damage to the vessel due to neutron irradiation, is essential for reasons of integrity of this enclosure and safety of any nuclear power plant. The objective of this thesis work is theoretical and experimentally determine the microstructural damage of a type nuclear reactor vessel steel BWR, due to neutron radiation from the reactor core, using microscopic and spectroscopic techniques as well as Monte Carlo simulation. Microscopy Optical, Scanning Electron Microscopy, Transmission Electron Microscopy, Energy Dispersion of X-rays Spectrometry and X-rays Diffractometry were the techniques used in this research. These techniques helped in the characterization of both the basis of design of pressure vessel steel and steel irradiated, after eight years of neutron irradiation on the vessel, allowing know the surface morphology and crystal structures of the previous steel and post-irradiation, analyze the change in the microstructure of the steel vessel, morphological damage to surface level in an irradiated sample, among which are cavities in the order of microns produced by Atomic displacements due to the impact of neutronic, above all in the first layers of thickness of the vessel, the effect of swelling, regions of greater damage and Atomic

  8. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects; Mecanismes de fragilisation sous irradiation aux neutrons d'alliages modeles ferritiques et d'un acier de cuve: amas de defauts

    Energy Technology Data Exchange (ETDEWEB)

    Meslin-Chiffon, E

    2007-11-15

    The embrittlement of reactor pressure vessel (RPV) under irradiation is partly due to the formation of point defects (PD) and solute clusters. The aim of this work was to gain more insight into the formation mechanisms of solute clusters in low copper ([Cu] = 0.1 wt%) FeCu and FeCuMnNi model alloys, in a copper free FeMnNi model alloy and in a low copper French RPV steel (16MND5). These materials were neutron-irradiated around 300 C in a test reactor. Solute clusters were characterized by tomographic atom probe whereas PD clusters were simulated with a rate theory numerical code calibrated under cascade damage conditions using transmission electron microscopy analysis. The confrontation between experiments and simulation reveals that a heterogeneous irradiation-induced solute precipitation/segregation probably occurs on PD clusters. (author)

  9. Neutron Irradiation Tests of Pressure Transducers in Liquid Helium

    CERN Document Server

    Amand, J F; Casas-Cubillos, J; Thermeau, J P

    1999-01-01

    The superconducting magnets of the future Large Hadron Collider (LHC) at CERN will operate in pressurised superfluid helium (1 bar, 1.9 K). About 500 pressure transducers will be placed in the liquid helium bath for monitoring the filling and the pressure transients after resistive transitions. Their precision must remain better than 100 mbar at pressures below 2 bar and better than 5% for higher pressures (up to 20 bar), with temperatures ranging from 1.8 K to 300 K. All the tested transducers are based on the same principle: the fluid or gas is separated from a sealed reference vacuum by an elastic membrane; its deformation indicates the pressure. The transducers will be exposed to high neutron fluence (2 kGy, 1014 n/cm2 per year) during the 20 years of machine operation. This irradiation may induce changes both on the membranes characteristics (leakage, modification of elasticity) and on gauges which measure their deformations. To investigate these effects and select the transducer to be used in the LHC, a...

  10. Response of neutron-irradiated RPV steels to thermal annealing

    Energy Technology Data Exchange (ETDEWEB)

    Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

    1997-03-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the fracture toughness properties that have been degraded by neutron irradiation. This paper summarizes experimental results of work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response of several irradiated RPV steels.

  11. The long-life design analysis of reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhang Ping

    2014-01-01

    Based on the fast neutron irradiation and thermal fatigue on reactor vessel ageing, the long-life design methods of reactor pressure vessel in EPR, AP1000, and CPR1000 are introduced. It is considered that adopting heavy reflector, increasing the downcomer of water gap, in-out refueling and increasing the toughness of the material are the methods to achieve the long-life running of RPV. (author)

  12. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

    Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

  13. Nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    McDonald, B.N.

    1976-01-01

    In nuclear power reactor systems which have a reactor core inside a pressure vessel, the feedwater inlet pipe and steam discharge nozzle usually require separate pressure vessel penetrations. This requirement involves a great deal of expensive high quality special machining, welding and weld joint testing. The invention overcomes most of these problems by nestling the feedwater inlet pipe inside the steam discharge nozzle. At the same time the individual heat exchanger modules are supported from the pressure vessel at the same location as the nested feedwater inlet pipe and steam discharge nozzle combination, thus eliminating the need to accomodate troublesome differential thermal expansion problems through special structures within the pressure vessel

  14. Modeling irradiation embrittlement in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

  15. Effects of irradiation at lower temperature on the microstructure of Cr-Mo-V-alloyed reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grosse, M.; Boehmert, J.; Gilles, R. [Hahn-Meitner-Institut Berlin GmbH (Germany)

    1998-10-01

    The microstructural damage process due to neutron irradiation [1] proceeds in two stages: - formation of displacement cascades - evolution of the microstructure by defect reactions. Continuing our systematic investigation about the microstructural changes of Russian reactor pressure vessel steel due to neutron irradiation the microstructure of two laboratory heats of the VVER 440-type reactor pressure vessel steel after irradiation at 60 C was studied by small angle neutron scattering (SANS). 60 C-irradiation differently changes the irradiation-induced microstructure in comparison with irradiation at reactor operation temperature and can, thus, provide new insights into the mechanisms of the irradiation damage. (orig.)

  16. Radiation embrittlement predictions for reactor pressure vessels

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Marston, T.U.

    1979-01-01

    The purpose of a surveillance program in a nuclear reactor pressure vessel is to provide an accurate measure of the changes in properties necessary to assess the structural integrity of the beltline region of the vessel as it is subjected to neutron irradiation damage. To understand problems surrounding a surveillance program, it is necessary to look at the concept of a structural integrity analysis. The structural integrity analysis can, of course, be only as accurate as the input data allows. Any errors in materials properties due to either errors in test technique to obtain these properties or to test specimens being exposed to environments different than actual operating condition will be reflected as errors in the actual structural integrity analysis. In a surveillance program, we must then be sure that we are measuring the proper material property to input to the fracture mechanics model as well as insure that the irradiation environment to which we subject the specimens will not cause an error in the prediction of the actual integrity margin of the pressure vessel beltline. The following text will question the appropriateness of using irradiated materials property data derived from high flux experiment, either in test reactors or advanced surveillance positions, to predict the actual material condition at the reactor vessel wall. Additionally, data will be presented showing several new correlation methods for deriving fracture toughness data from Charpy specimens, thus giving the proper input to the fracture mechanics model of a structural integrity analysis

  17. PRESSURE-RESISTANT VESSEL

    NARCIS (Netherlands)

    Beukers, A.; De Jong, T.

    1997-01-01

    Abstract of WO 9717570 (A1) The invention is directed to a wheel-shaped pressure-resistant vessel for gaseous, liquid or liquefied material having a substantially rigid shape, said vessel comprising a substantially continuous shell of a fiber-reinforced resin having a central opening, an inner

  18. Reactor pressure vessel support

    International Nuclear Information System (INIS)

    Butti, J.P.

    1978-01-01

    A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction

  19. Pressurized Vessel Slurry Pumping

    International Nuclear Information System (INIS)

    Pound, C.R.

    2001-01-01

    This report summarizes testing of an alternate ''pressurized vessel slurry pumping'' apparatus. The principle is similar to rural domestic water systems and ''acid eggs'' used in chemical laboratories in that material is extruded by displacement with compressed air

  20. Sapphire tube pressure vessel

    Science.gov (United States)

    Outwater, John O.

    2000-01-01

    A pressure vessel is provided for observing corrosive fluids at high temperatures and pressures. A transparent Teflon bag contains the corrosive fluid and provides an inert barrier. The Teflon bag is placed within a sapphire tube, which forms a pressure boundary. The tube is received within a pipe including a viewing window. The combination of the Teflon bag, sapphire tube and pipe provides a strong and inert pressure vessel. In an alternative embodiment, tie rods connect together compression fittings at opposite ends of the sapphire tube.

  1. Pressure vessel, in particular reactor pressure vessel

    International Nuclear Information System (INIS)

    1976-01-01

    The task to design the pre-stressing facility for a pressure vessel, especially for nuclear reactors, wiht pre-stressed jacket and pre-stressing facility, the latter one showing circumferential steel tendons supported polygonally on the outer side of the jacket by means of supporting shoes, in such a way that simply defined tensions can be generated and re-tensioning of the tendons can be carried out easily is solved according to the invention by keeping the circumference of the steel tendons fixed and by designing the supporting shoes as stressing shoes. The defined tensions are applied through the stressing shoes. (ORU) [de

  2. Attachment Fitting for Pressure Vessel

    Science.gov (United States)

    Smeltzer, Stanley S., III (Inventor); Carrigan, Robert W. (Inventor)

    2002-01-01

    This invention provides sealed access to the interior of a pressure vessel and consists of a tube. a collar, redundant seals, and a port. The port allows the seals to be pressurized and seated before the pressure vessel becomes pressurized.

  3. GOLD PRESSURE VESSEL SEAL

    Science.gov (United States)

    Smith, A.E.

    1963-11-26

    An improved seal between the piston and die member of a piston-cylinder type pressure vessel is presented. A layer of gold, of sufficient thickness to provide an interference fit between the piston and die member, is plated on the contacting surface of at least one of the members. (AEC)

  4. Pressure vessel integrity 1991

    International Nuclear Information System (INIS)

    Bhandari, S.; Doney, R.O.; McDonald, M.S.; Jones, D.P.; Wilson, W.K.; Pennell, W.E.

    1991-01-01

    This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion

  5. Revisiting the reactor pressure vessel for long-time operation

    International Nuclear Information System (INIS)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-01-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  6. Reactor pressure vessel nozzle

    Science.gov (United States)

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

  7. Quenchable compressed graphite synthesized from neutron-irradiated highly oriented pyrolytic graphite in high pressure treatment at 1500 °C

    Science.gov (United States)

    Niwase, Keisuke; Terasawa, Mititaka; Honda, Shin-ichi; Niibe, Masahito; Hisakuni, Tomohiko; Iwata, Tadao; Higo, Yuji; Hirai, Takeshi; Shinmei, Toru; Ohfuji, Hiroaki; Irifune, Tetsuo

    2018-04-01

    The super hard material of "compressed graphite" (CG) has been reported to be formed under compression of graphite at room temperature. However, it returns to graphite under decompression. Neutron-irradiated graphite, on the other hand, is a unique material for the synthesis of a new carbon phase, as reported by the formation of an amorphous diamond by shock compression. Here, we investigate the change of structure of highly oriented pyrolytic graphite (HOPG) irradiated with neutrons to a fluence of 1.4 × 1024 n/m2 under static pressure. The neutron-irradiated HOPG sample was compressed to 15 GPa at room temperature and then the temperature was increased up to 1500 °C. X-ray diffraction, high-resolution transmission electron microscopy on the recovered sample clearly showed the formation of a significant amount of quenchable-CG with ordinary graphite. Formation of hexagonal and cubic diamonds was also confirmed. The effect of irradiation-induced defects on the synthesis of quenchable-CG under high pressure and high temperature treatment was discussed.

  8. Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

    1998-07-01

    SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

  9. High pressure storage vessel

    Science.gov (United States)

    Liu, Qiang

    2013-08-27

    Disclosed herein is a composite pressure vessel with a liner having a polar boss and a blind boss a shell is formed around the liner via one or more filament wrappings continuously disposed around at least a substantial portion of the liner assembly combined the liner and filament wrapping have a support profile. To reduce susceptible to rupture a locally disposed filament fiber is added.

  10. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  11. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    International Nuclear Information System (INIS)

    Wang, Jy-An John

    2010-01-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

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

    International Nuclear Information System (INIS)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

    1999-01-01

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

  13. Radiation effects on reactor pressure vessel supports

    International Nuclear Information System (INIS)

    Johnson, R.E.

    1996-05-01

    The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

  14. Comparison of four NDT methods for indication of reactor steel degradation by high fluences of neutron irradiation

    Czech Academy of Sciences Publication Activity Database

    Tomáš, Ivan; Vértesy, G.; Pirfo Barroso, S.; Kobayashi, S.

    2013-01-01

    Roč. 265, DEC (2013), s. 201-209 ISSN 0029-5493 Institutional support: RVO:68378271 Keywords : neutron irradiation * steel degradation * nuclear reactor pressure vessel * magnetic NDT * magnetic minor hysteresis loops * Magnetic Barkhausen Emission Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 0.972, year: 2013 http://www.sciencedirect.com/science/article/pii/S0029549313004664

  15. Lateral expansion and impact toughness correlation of VVER-1000 reactor pressure vessel materials

    Directory of Open Access Journals (Sweden)

    O. V. Trygubenko

    2014-12-01

    Full Text Available Impact toughness and lateral expansion of the irradiated surveillance specimens for VVER-1000 reactor pres-sure vessel have been defined using Charpy impact test. The analysis of experimental data has revealed a linear correlation of these characteristics. It was shown that a slope of the regression line is practically unchanged due to irradiation. Using the upper shelf energy test data it was also demonstrated the lateral expansion and impact toughness decrease simultaneously under the neutron irradiation.

  16. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel

    International Nuclear Information System (INIS)

    Moranchel y Rodriguez, M.; Garcia B, A.; Longoria G, L. C.

    2010-09-01

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  17. Ageing Management Review of the reactor pressure vessels in Laguna Verde NPP

    International Nuclear Information System (INIS)

    Gris Cruz, Magdalena; Arganis, Carlos R.J.; Medina Almazan, A. Liliana

    2012-01-01

    In the present paper, for both units of Laguna Verde Nuclear Power Plant (LVNPP), the Ageing Management Review of the reactor pressure Vessel was carried out, including the identification of the intended functions, the materials and the environments. The evaluation of the ageing effect/mechanism and the Aging management programs currently implemented were prepared. The most important aging effects/ mechanisms are: loss of fracture toughness due to neutron irradiation embrittlement, fatigue, stress corrosion cracking (SCC), general corrosion and erosion-corrosion. The neutron irradiation embrittlement is managed by the reactor vessel materials surveillance program. The fatigue is a Time Limited Aging Analysis (TLAA), for which is necessary to calculate some fatigue usage factors. SCC is managed by, the In service inspections (ISI) program, but also by the Water Chemistry program, including, currently, On Line Noble Chem. The water chemistry program also manages General Corrosion and erosion-corrosion. The results were compared with the GALL report. (author)

  18. Identification of neutron irradiation induced strain rate sensitivity change using inverse FEM analysis of Charpy test

    Science.gov (United States)

    Haušild, Petr; Materna, Aleš; Kytka, Miloš

    2015-04-01

    A simple methodology how to obtain additional information about the mechanical behaviour of neutron-irradiated WWER 440 reactor pressure vessel steel was developed. Using inverse identification, the instrumented Charpy test data records were compared with the finite element computations in order to estimate the strain rate sensitivity of 15Ch2MFA steel irradiated with different neutron fluences. The results are interpreted in terms of activation volume change.

  19. Identification of neutron irradiation induced strain rate sensitivity change using inverse FEM analysis of Charpy test

    Energy Technology Data Exchange (ETDEWEB)

    Haušild, Petr, E-mail: petr.hausild@fjfi.cvut.cz [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Materials, Trojanova 13, 120 00 Praha 2 (Czech Republic); Materna, Aleš [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Materials, Trojanova 13, 120 00 Praha 2 (Czech Republic); Kytka, Miloš [Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering, Department of Materials, Trojanova 13, 120 00 Praha 2 (Czech Republic); Nuclear Research Institut, ÚJV Řež, a.s., Hlavní 130, Řež, 250 68 Husinec (Czech Republic)

    2015-04-15

    A simple methodology how to obtain additional information about the mechanical behaviour of neutron-irradiated WWER 440 reactor pressure vessel steel was developed. Using inverse identification, the instrumented Charpy test data records were compared with the finite element computations in order to estimate the strain rate sensitivity of 15Ch2MFA steel irradiated with different neutron fluences. The results are interpreted in terms of activation volume change.

  20. Reactor pressure vessel embrittlement

    International Nuclear Information System (INIS)

    1992-07-01

    Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

  1. 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)

  2. Moessbauer effect on a reactor pressure vessel irradiated with fast neutrons

    International Nuclear Information System (INIS)

    Yi, Y. G.; Kim, H. S.; Hong, C. Y.; Jang, K. S.; Yoo, Y. B.

    1998-01-01

    To estimate the integrity of a nuclear-reactor pressure vessel (RPV) subjected to neutron irradiation, we studied the changes in the material properties of RPV steel irradiated by fast neutrons (E n > 1 MeV) at a level of 0 ∼ 10 18 n/cm 2 . An X-ray diffraction experiment was done to investigate the changes in the phases of all specimens, nd the influence of the neutron irradiation on the RPV steel studied by the Mossbauer effect. The absorption area decreased rapidly for doses above 10 17 n/cm 2 , but at site 3 the value of the isomer shift and the quadrupole splitting increased. The magnetic hyperfine field at site 1 and 2 varied only within ±3 %

  3. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

  4. Neutron irradiation therapy machine

    International Nuclear Information System (INIS)

    1980-01-01

    Conventional neutron irradiation therapy machines, based on the use of cyclotrons for producing neutron beams, use a superconducting magnet for the cyclotron's magnetic field. This necessitates complex liquid He equipment and presents problems in general hospital use. If conventional magnets are used, the weight of the magnet poles considerably complicates the design of the rotating gantry. Such a therapy machine, gantry and target facilities are described in detail. The use of protons and deuterons to produce the neutron beams is compared and contrasted. (U.K.)

  5. Ductile fracture estimation of reactor pressure vessel under thermal shock

    International Nuclear Information System (INIS)

    Takahashi, Jun; Sakai, Shinsuke; Okamura, Hiroyuki

    1990-01-01

    This paper presents a new scheme for the estimation of unstable ductile fracture of a reactor pressure vessel under thermal shock conditions. First, it is shown that the bending moment applied to the cracked section can be evaluated by considering the plastic deformation of the cracked section and the thermal deformation of the shell. As the contribution of the local thermal stress to the J-value is negligible, the J-value under thermal shock can be easily evaluated by using fully plastic solutions for the cracked part. Next, the phenomena of ductile fracture under thermal shock are expressed on the load-versus-displacement diagram which enables us to grasp the transient phenomena visually. In addition, several parametrical surveys are performed on the above diagram concerning the variation of (1) thermal shock conditions, (2) initial crack length, and (3) J-resistance curve (i.e. embrittlement by neutron irradiation). (author)

  6. Technical note: irradiation embrittlement of pressure vessel steels, analysis of the IAEA coordinated program results

    International Nuclear Information System (INIS)

    Hammad, F.H.; Ghoneim, M.M.; Abou-Zahra, A.

    1985-01-01

    The embrittlement of certain steels as the result of neutron irradiation has significance for evaluating the risks associated with pressurized thermal shock and other possible pressure vessel failure mechanisms, especially in the heat-affected zones of welds of older pressure vessels. A knowledge of the degree of embrittlement associated with a given integral fluence, usually expressed as an increase in the ductile-brittle transition temperature, is therefore needed to estimate the service life of pressure vessels subject to such embrittlement. This Technical Note describes a reanalysis of the results of the International Atomic Energy Agency coordinated program to measure this effect, which succeeded in explaining and reducing the very large degree of scatter in the results originally obtained in the measurements

  7. Special enclosure for a pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.; Wedellsborg, U.W.

    1993-01-01

    A pressure vessel enclosure is described comprising a primary pressure vessel, a first pressure vessel containment assembly adapted to enclose said primary pressure vessel and be spaced apart therefrom, a first upper pressure vessel jacket adapted to enclose the upper half of said first pressure vessel containment assembly and be spaced apart therefrom, said upper pressure vessel jacket having an upper rim and a lower rim, each of said rims connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, mean for connecting in a sealable relationship said upper rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, means for connecting in a sealable relationship said lower rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a first lower pressure vessel jacket adapted to enclose the lower half of said first pressure vessel containment assembly and be spaced apart therefrom, said lower pressure vessel jacket having an upper rim connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, and means for connecting in a sealable relationship said upper rim of said first lower pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a second upper pressure vessel jacket adapted to enclose said first upper pressure vessel jacket and be spaced apart therefrom, said second upper pressure vessel jacket having an upper rim and a lower rim, each of said rims adapted to slidably engage the outer surface of said first upper pressure vessel jacket, means for sealing said rims, a second lower pressure vessel jacket adapted to enclose said first lower pressure vessel jacket and be spaced apart therefrom

  8. Level indicator for pressure vessels

    Science.gov (United States)

    Not Available

    1982-04-28

    A liquid-level monitor for tracking the level of a coal slurry in a high-pressure vessel including a toroidal-shaped float with magnetically permeable bands thereon disposed within the vessel, two pairs of magnetic-field generators and detectors disposed outside the vessel adjacent the top and bottom thereof and magnetically coupled to the magnetically permeable bands on the float, and signal-processing circuitry for combining signals from the top and bottom detectors for generating a monotonically increasing analog control signal which is a function of liquid level. The control signal may be utilized to operate high-pressure control valves associated with processes in which the high-pressure vessel is used.

  9. Description of code system PLES/PTS for evaluation of pressure vessel integrity during PTS events

    International Nuclear Information System (INIS)

    Hirano, Masashi; Kohsaka, Atsuo.

    1992-02-01

    A code system PLES/PTS has been developed at the Japan Atomic Energy Research Institute (JAERI) to evaluate the integrity of the pressure vessel during plant thermal-hydraulic transients related to pressurized thermal shock (PTS) in a pressurized water reactor (PWR). The code system consists of several member codes to analyse the thermal-mixing behavior of emergency core cooling (ECC) water and primary coolant, transient stress distribution within the vessel wall, and crack growth behavior at the inner surface of the vessel. The crack growth behavior is evaluated by comparing the stress intensity factor (k I ) with the crack initiation toughness (k Ic ) and crack arrest toughness (k Ic ), taking into account the fast neutron irradiation embrittlement. This report describes the methods and models applied in PLES/PTS and the input data requirements. (author)

  10. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Stoll, A.

    1976-01-01

    The invention relates to a pressure vessel which can be used for nuclear reactors and for chemical processing technologies. A grid of walls welded to each other, which is installed in the interior of the pressure vessel, is so attached to an outer jacket at several edges, that these edges exert a force on the wall of the vessel directed towards the interior. Only the out jacket resists the differential between the inner and outer pressures; the welded walls in the interior do not have to sustain any differential pressure. They create a larger number of inner spaces (or tubes) which can be individually accessible and each of which has a terminal element. (UWI) [de

  11. Diamond amorphization in neutron irradiation

    International Nuclear Information System (INIS)

    Nikolaenko, V.A.; Gordeev, V.G.

    1996-01-01

    The paper presents the results on neutron irradiation of the diamond in a nuclear reactor. It is shown that the neutron irradiation stimulates the diamond transition to the amorphous state. At a temperature below 750 o K the time required for the diamond-graphite transition decreases with decreasing irradiation temperature. On the contrary, in irradiation at higher temperatures the time of diamond conversion into the amorphous state increases with decreasing but always remains shorter than in the absence of irradiation. (author)

  12. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

    Strosnider, J.; Wichman, K.; Elliot, B.

    1994-12-01

    This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

  13. Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

    Czech Academy of Sciences Publication Activity Database

    Tomáš, Ivan; Vértesy, G.; Gillemot, F.; Székely, R.

    2012-01-01

    Roč. 432, 1-3 (2012), s. 371-377 ISSN 0022-3115 R&D Projects: GA ČR GA101/09/1323 Institutional research plan: CEZ:AV0Z10100520 Keywords : neutron irradiation * steel degradation * nuclear reactor pressure vessel * magnetic NDT * magnetic minor hysteresis loops * Magnetic Adaptive Testing Subject RIV: BM - Solid Matter Physics ; Magnetism Impact factor: 1.211, year: 2012 http://dx.doi.org/10.1016/j.jnucmat.2012.09.006

  14. 46 CFR 119.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels. 119.330 Section 119.330 Shipping COAST... Machinery § 119.330 Pressure vessels. All unfired pressure vessels must be installed to the satisfaction of the cognizant OCMI. The design, construction, and original testing of such unfired pressure vessels...

  15. 46 CFR 182.330 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels. 182.330 Section 182.330 Shipping COAST...) MACHINERY INSTALLATION Auxiliary Machinery § 182.330 Pressure vessels. All unfired pressure vessels must be... unfired pressure vessels must meet the applicable requirements of subchapter F (Marine Engineering) of...

  16. 46 CFR 169.249 - Pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels. 169.249 Section 169.249 Shipping COAST... and Certification Inspections § 169.249 Pressure vessels. Pressure vessels must meet the requirements of part 54 of this chapter. The inspection procedures for pressure vessels are contained in subpart...

  17. Stress analysis of pressure vessels

    International Nuclear Information System (INIS)

    Kim, B.K.; Song, D.H.; Son, K.H.; Kim, K.S.; Park, K.B.; Song, H.K.; So, J.Y.

    1979-01-01

    This interim report contains the results of the effort to establish the stress report preparation capability under the research project ''Stress analysis of pressure vessels.'' 1978 was the first year in this effort to lay the foundation through the acquisition of SAP V structural analysis code and a graphic terminal system for improved efficiency of using such code. Software programming work was developed in pre- and post processing, such as graphic presentation of input FEM mesh geometry and output deformation or mode shope patterns, which was proven to be useful when using the FEM computer code. Also, a scheme to apply fracture mechanics concept was developed in fatigue analysis of pressure vessels. (author)

  18. Pressure vessel and method therefor

    Science.gov (United States)

    Saunders, Timothy

    2017-09-05

    A pressure vessel includes a pump having a passage that extends between an inlet and an outlet. A duct at the pump outlet includes at least one dimension that is adjustable to facilitate forming a dynamic seal that limits backflow of gas through the passage.

  19. Pressure vessel having continuous sidewall

    Science.gov (United States)

    Simon, Xavier D. (Inventor); Barackman, Victor J. (Inventor)

    2011-01-01

    A spacecraft pressure vessel has a tub member. A sidewall member is coupled to the tub member so that a bottom section of the sidewall member extends from an attachment intersection with the tub member and away from the tub member. The bottom section of the sidewall member receives and transfers a load through the sidewall member.

  20. The reactor vessel steels

    International Nuclear Information System (INIS)

    Bilous, W.; Hajewska, E.; Szteke, W.; Przyborska, M.; Wasiak, J.; Wieczorkowski, M.

    2005-01-01

    In the paper the fundamental steels using in the construction of pressure vessel water reactor are discussed. The properties of these steels as well as the influence of neutron irradiation on its degradation in the time of exploitation are also done. (authors)

  1. JAIPEC's R and D activity on neutron irradiation embrittlement of RPV materials

    International Nuclear Information System (INIS)

    Otsuka, T.

    1999-01-01

    Japan Power Engineering and Inspection Corporation (JAPEIC) is managing many national RD projects in the field related to inspection, welding and integrity evaluation according to the policy of the Ministry of International Trade and Industry (MITI) on the safety and reliability of operating nuclear power plants. The Plant Life Management Technology Project (PLIM), one of these RD projects, has newly started from 1996 to develop the evaluation methodologies for the neutron irradiation embrittlement in the upper shelf region of the reactor pressure vessel (RPV) and the thermal embrittlement of duplex stainless steel components, and to develop the technology for reconstitution of RPV surveillance test pieces. For the first item of these three, the development of evaluation method in the upper shelf region, the detailed scope of test program has been recently settled and the neutron irradiation of test specimens is scheduled to start this year using the Halden Boiling Water Reactor (HBWR) in Norway. This paper presents the RD scope for the development of evaluation method for the neutron irradiation embrittlement in the upper shelf region in the PLIM project of JAPEIC (author) (ml)

  2. 46 CFR 197.462 - Pressure vessels and pressure piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and pressure piping. 197.462 Section... Diving Equipment § 197.462 Pressure vessels and pressure piping. (a) The diving supervisor shall ensure that each pressure vessel, including each volume tank, cylinder and PVHO, and each pressure piping...

  3. Use of probability with linear elastic fracture mechanics in studying brittle fracture in pressure vessels

    International Nuclear Information System (INIS)

    Jouris, G.M.; Shaffer, D.H.

    1978-01-01

    One phase is considered in the development of statistical methodology for a fracture mechanics analysis of the failure of nuclear steam supply system components, in particular that of brittle fracture in the beltline region of the pressure vessel resulting from various transients. It introduces a probability structure into the deterministic linear elastic fracture mechanics calculations. The resulting estimates of probability of brittle fracture reflect not only variation due to heterogeneity of vessel material but also uncertainties in the effect of embrittlement of the vessel steel due to neutron irradiation. Using importance sampling in conjunction with Monte Carlo simulation it was estimated that, for the operational transients considered, the probability of brittle fracture conditional on the presence of an Appendix G (ASME Code.Sn13) flaw is less than 2 x 10sup(-10). (author)

  4. Neutron irradiation effects on plasma facing materials

    Science.gov (United States)

    Barabash, V.; Federici, G.; Rödig, M.; Snead, L. L.; Wu, C. H.

    2000-12-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed.

  5. Neutron irradiation effects on plasma facing materials

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Roedig, M.; Snead, L.L.; Wu, C.H.

    2000-01-01

    This paper reviews the effects of neutron irradiation on thermal and mechanical properties and bulk tritium retention of armour materials (beryllium, tungsten and carbon). For each material, the main properties affected by neutron irradiation are described and the specific tests of neutron irradiated armour materials under thermal shock and disruption conditions are summarized. Based on current knowledge, the expected thermal and structural performance of neutron irradiated armour materials in the ITER plasma facing components are analysed

  6. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Pareige, P.

    1996-04-01

    Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends

  7. Prevention of non-ductile fracture in 6061-T6 aluminum nuclear pressure vessels

    International Nuclear Information System (INIS)

    Yahr, G.T.

    1995-01-01

    The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Committee has approved rules for the use of 6061-T6 and 6061-T651 aluminum for the construction of Class 1 welded nuclear pressure vessels for temperatures not exceeding 149 C (300 F). Nuclear Code Case N-519 allows the use of this aluminum in the construction of low temperature research reactors such as the Advanced Neutron Source. The rules for protection against non-ductile fracture are discussed. The basis for a value of 25.3 MPa √m (23 ksi √in.) for the critical or reference stress intensity factor for use in the fracture analysis is presented. Requirements for consideration of the effects of neutron irradiation on the fracture toughness are discussed

  8. Manufacturing technologies of PWR pressure vessels

    International Nuclear Information System (INIS)

    Qin Xubin

    1991-01-01

    Pressure vessels belong to the main component of PWR plants. Starting with describing the manufacture of pressure vessel components and their assembly, the manufacturing technologies of pressure vessels are briefly presented with regards to welding, heat treatment, inspections and testing. In addition, quality assurance during the manufacture is presented with emphasis

  9. Model tests for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Stoever, R.

    1975-01-01

    Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

  10. Reactor pressure vessel vented head

    Science.gov (United States)

    Sawabe, James K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell.

  11. Application of advanced irradiation analysis methods to light water reactor pressure vessel test and surveillance programs

    International Nuclear Information System (INIS)

    Odette, R.; Dudey, N.; McElroy, W.; Wullaert, R.; Fabry, A.

    1977-01-01

    Inaccurate characterization and inappropriate application of neutron irradiation exposure variables contribute a substantial amount of uncertainty to embrittlement analysis of light water reactor pressure vessels. Damage analysis involves characterization of the irradiation environment (dosimetry), correlation of test and surveillance metallurgical and dosimetry data, and projection of such data to service conditions. Errors in available test and surveillance dosimetry data are estimated to contribute a factor of approximately 2 to the data scatter. Non-physical (empirical) correlation procedures and the need to extrapolate to the vessel may add further error. Substantial reductions in these uncertainties in future programs can be obtained from a more complete application of available damage analysis tools which have been developed for the fast reactor program. An approach to reducing embrittlement analysis errors is described, and specific examples of potential applications are given. The approach is based on damage analysis techniques validated and calibrated in benchmark environments

  12. Evaluation of the reactor pressure vessel steels by positron annihilation

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, V., E-mail: Vladimir.Slugen@elf.stuba.sk [Institute of Nuclear and Physical Engineering, Slovak University of Technology, Ilkovičova 3, 81219 Bratislava (Slovakia); Hein, H. [AREVA NP GmbH, Paul Gossen Strasse 100, 91 001 Erlangen (Germany); Sojak, S.; Simeg Veterníková, J.; Petriska, M.; Sabelová, V.; Pavúk, M.; Hinca, R.; Stacho, M. [Institute of Nuclear and Physical Engineering, Slovak University of Technology, Ilkovičova 3, 81219 Bratislava (Slovakia)

    2013-11-15

    This paper presents a comparison of commercially used German and Russian reactor pressure vessel steels from the positron annihilation spectroscopy (PAS) point of view, having in mind knowledge obtained also from other techniques from the last decades. The second generation of Russian RPV steels seems to be fully comparable with German steels and their quality allows prolongation of NPP operating lifetime over projected 40 years. The embrittlement of CrMoV steels is relatively low due to effect of higher temperature which implies partial in situ annealing of primary microstructural point defects and therefore delays the degradation processes caused by neutron irradiation. PAS techniques can be effectively applied for evaluation of microstructural changes caused by extreme external loads (characterized by high dpa values) by proton implantation, with aim to simulate irradiation and for the evaluation of the effectiveness of post-irradiation thermal treatments. We used our actual and previous results, collected during last 20 years from measurements of different RPV-steels in “as received”, irradiated and post-irradiation annealed state and compare them with the aim to contribute to general knowledge based on experimental PAS data. Actual results from irradiated German and Russian steels confirmed that no large voids or vacancy clusters were formed at defined irradiation conditions stated according to the real operational conditions at nuclear power plants. This indicate the fact that vacancy type defects bear hardly any responsibility for radiation-induced hardening and embrittlement of reactor pressure vessel steels and does not affect significantly the long-term operation of nuclear power plants from safety point of view.

  13. Stabilizer for reactor pressure vessel

    International Nuclear Information System (INIS)

    Nakajima, Masataka.

    1991-01-01

    A plurality of flush springs and sleeves are disposed to a rod to be screw-coupled to a yoke disposed between gusset side walls. A vibration energy dissipation mechanism due to resiliently plastic deformation of metal is disposed between the gusset side wall and the sleeve of the rod. When earthquakes should occur and the distance between the gusset side wall and the sleeve is changed undergoing seismic input, the vibration energy is dissipated by the resilient plastic deformation of the vibration energy dissipation mechanism. The seismic response of the pressure vessel system is thus decreased to improve seismic resistivity. The reactor safety and reliability can be ensured. It is not necessary to strengthen supports for pipeline systems, and increase of the cost, including design and manufacture, can be avoided while giving no effects on the layout for the upper portion of reactor shielding walls. (N.H.)

  14. Postirradiation recovery of a reactor pressure vessel steel investigated by positron annihilation and microhardness measurements

    International Nuclear Information System (INIS)

    Pareja, R.; Diego, N. De; Cruz, R.M. de la; Del Rio, J.

    1993-01-01

    Positron lifetime and microhardness measurements have been performed on untreated, thermal-aged, neutron-irradiated, and postirradiation-annealed samples of reactor pressure vessel steels with the purpose of investigating the mechanisms of irradiation-induced hardening and recovery of the mechanical properties in these materials. The positron lifetime experiments have not revealed any evidence of the formation of a significant concentration of voids or vacancy clusters in samples irradiated at ∼290 C with fluences ≤2.71 x 10 23 n/m 2 (E>1 MeV), but they suggest a dislocation annealing induced by the irradiation. Isochronal annealing experiments with neutron-irradiated samples show a simultaneous recovery in their positron lifetime and microhardness at ∼340 C. From the microhardness measurements, the yield strength of the irradiated material has been estimated. The results appear to be consistent with a model of hardening due to irradiation-induced dissolution of precipitates with formation of small metastable precipitates after postirradiation aging and recovery induced by the disappearance of these metastable precipitates

  15. An assessment of the failure rate for the beltline region of PWR pressure vessels during normal operation and certain transient conditions. Technical report

    International Nuclear Information System (INIS)

    Gamble, R.M.; Strosnider, J. Jr.

    1981-06-01

    A study was conducted to assess the failure rate for the beltline region of a generic pressurized-water reactor (PWR) pressure vessel. This assessment included the evaluation of several normal operating and transient reactor conditions. Failure rates were calculated from a computer code that used fracture mechanics methods to model the failure process; random number generation techniques were used to simulate random variables and model their interaction in the failure-process. This investigation had three major objectives: (1) to better define the effect of neutron irradiation, material variation, and flaw distribution on the failure rate for the beltline region of PWR pressure vessels, (2) to estimate the relative margins against failure for normal operation and certain transient conditions associated with nuclear pressure vessels, and (3) to evaluate the current limitations for using fracture mechanics models to predict failure rates for nuclear pressure vessels

  16. Post-irradiation fatigue test of 2 1/4Cr-1Mo steel for the pressure vessel material of the HTTR

    International Nuclear Information System (INIS)

    Ishii, Toshimitsu; Fukaya, Kiyoshi; Nishiyama, Yutaka; Suzuki, Masahide; Eto, Motokuni; Ohmi, Masao; Mimura, Hideaki; Ooka, Norikazu.

    1996-06-01

    Low cycle fatigue tests of the irradiated 2 1/4Cr-1Mo steel for the pressure vessel material of the High Temperature Engineering Test Reactor (HTTR) were performed to examine the neutron irradiation effect on fatigue properties, i.e., fatigue life and cyclic softening behavior. These tests were performed at 450degC in a vacuum of ∼10 -4 Pa using a servo-hydraulic fatigue testing machine installed in a hot cell of the Hot Laboratory of the Oarai Research Establishment. Axial strain rate was controlled at 0.1%/sec with total strain ranges of 0.75, 0.8, 1.0, 1.2 and 1.5%. Main results are summarized as follows: 1) Maximum stress at the initial stage of the cyclic fatigue testing of the specimens irradiated to a neutron fluence over 2x10 19 n/cm 2 (>1MeV) increased in comparison with the unirradiated specimens. 2) No significant difference was found in the cyclic softening behavior between the irradiated and unirradiated specimens. 3) Fatigue life of the irradiated specimen whose ductility was reduced by the neutron irradiation decreased slightly in comparison with the unirradiated specimen. 4) It is considered that there is no effect of the neutron irradiation on the fatigue life and the cyclic softening behavior at the total fluence expected in the lifetime of the HTTR pressure vessel. (author)

  17. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  18. Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

    International Nuclear Information System (INIS)

    Krasikov, E.A.; Nikolaenko, V.A.

    2008-01-01

    Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

  19. How to replace a reactor pressure vessel

    International Nuclear Information System (INIS)

    Huber, R.

    1996-01-01

    A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

  20. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  1. Lightweight bladder lined pressure vessels

    Science.gov (United States)

    Mitlitsky, Fred; Myers, Blake; Magnotta, Frank

    1998-01-01

    A lightweight, low permeability liner for graphite epoxy composite compressed gas storage vessels. The liner is composed of polymers that may or may not be coated with a thin layer of a low permeability material, such as silver, gold, or aluminum, deposited on a thin polymeric layer or substrate which is formed into a closed bladder using torispherical or near torispherical end caps, with or without bosses therein, about which a high strength to weight material, such as graphite epoxy composite shell, is formed to withstand the storage pressure forces. The polymeric substrate may be laminated on one or both sides with additional layers of polymeric film. The liner may be formed to a desired configuration using a dissolvable mandrel or by inflation techniques and the edges of the film seamed by heat sealing. The liner may be utilized in most any type of gas storage system, and is particularly applicable for hydrogen, gas mixtures, and oxygen used for vehicles, fuel cells or regenerative fuel cell applications, high altitude solar powered aircraft, hybrid energy storage/propulsion systems, and lunar/Mars space applications, and other applications requiring high cycle life.

  2. Mechanical spectroscopy of reactor-pressure-vessel steel embrittlement: a progress report

    International Nuclear Information System (INIS)

    Van Ouytsel, K.

    1998-08-01

    An enhanced surveillance strategy for testing the fracture toughness of reactor-pressure-vessel steel embrittlement is described. Microstructural investigation in support of damage modelling is an essential element in this enhanced strategy. Temperature-dependent experiments are very sensitive to differences in chemical composition and to effects of neutron irradiation as well as thermal ageing. Amplitude-dependent experiments can be related to tensile test results and correspond to a model for the yield stress. A full range of experiments were carried out on base and weld metal from the Doel-I-II power plants. The results have indicated that internal friction yields information which cannot always be detected by means of standard testing techniques. An inverted torsion pendulum for measuring internal friction has been constructed

  3. Mechanical spectroscopy of reactor-pressure-vessel steel embrittlement: a progress report

    Energy Technology Data Exchange (ETDEWEB)

    Van Ouytsel, K

    1998-08-01

    An enhanced surveillance strategy for testing the fracture toughness of reactor-pressure-vessel steel embrittlement is described. Microstructural investigation in support of damage modelling is an essential element in this enhanced strategy. Temperature-dependent experiments are very sensitive to differences in chemical composition and to effects of neutron irradiation as well as thermal ageing. Amplitude-dependent experiments can be related to tensile test results and correspond to a model for the yield stress. A full range of experiments were carried out on base and weld metal from the Doel-I-II power plants. The results have indicated that internal friction yields information which cannot always be detected by means of standard testing techniques. An inverted torsion pendulum for measuring internal friction has been constructed.

  4. The metrological problems of irradiation embrittlement of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Vodenicharov, S.; Kamenova, Ts.

    1993-01-01

    Neutron irradiation of reactor pressure vessel steels increases the T k -values of transition temperature from ductile to brittle fracture. This effect is very important in emergency situations, when the water cooling injection in the reactor results in high thermal gradients. In such cases there is a risk from the appearance of a brittle fracture with catastrophic crack propagation speed at relatively low stresses. That is why the T k -value determination is very important for the safe operation of the reactor systems. Some advanced experimental methods for T k -testing and control have been discussed in the present article and the standards of different countries have been compared. The methods applying subsize specimens and welding-restored specimens have been reviewed. (author)

  5. Dislocation-defect interactions in nuclear reactor pressure-vessel steels investigated by means of internal friction

    International Nuclear Information System (INIS)

    Van Ouytsel, K.; Fabry, A.; Schaller, R.

    1999-01-01

    Full text: A study of pressure-vessel steel embrittlement by means of temperature-dependent and amplitude-dependent internal friction has been carried out within the framework of commercial surveillance of nuclear reactor components. An inverted torsion pendulum operating at approximately 1 Hz has been employed to study JRQ, Doel-I-II and Doel IV pressure-vessel steels in unirradiated, thermally aged, neutron-irradiated, and annealed conditions. The temperature-dependent experiments evidence a reduction in the dislocation mobility as a result of neutron irradiation and prove that the technique is sensitive to thermal ageing and annealing involving changes in the dislocation mobility and type of dislocation-defect interaction. Amplitude-dependent results reveal a critical amplitude which can be related to the yield stress of the material in tension and corresponds well to a two-component model of the yield stress encompassing contributions of a short-range nature, such as the Peierls barrier to dislocation movement, and of long-range character, such as copper-rich precipitation and grain boundary segregation. The applicability of the non-destructive internal-friction technique to commercial steel programmes is highlighted. (author)

  6. The multifunction neutron irradiator (MNI)

    International Nuclear Information System (INIS)

    Zhou, Yongmao; Li, Shenzhi

    1994-01-01

    The Multifunction Neutron Irradiator (MNI) presented is a small-type neutron source reactor, for usage in the Boron Neutron Capture Therapy for human brain glioblastoma, Instrumental Neutron Activation Analysis (INAA), short-lived radioisotope production, and some fundamental researches. The reactor core is designed to have passive safety and the process control of the reactor operations is fully computerized. There are two operational modes: The routine operation mode with reactor power 20 ∼ 30 kW and flux 1 x 10 12 n · cm -2 · s -1 and the enhanced power operation mode for medical use. The irradiator can be located in a medical centre, research institute or university. 4 refs., 1 tab., 1 fig

  7. Emergency venting of pressure vessels

    International Nuclear Information System (INIS)

    Steinkamp, H.

    1995-01-01

    With the numerical codes developed for safety analysis the venting of steam vessel can be simulated. ATHLET especially is able to predict the void fraction depending on the vessel height. Although these codes contain a one-dimensional model they allow the description of complex geometries due to the detailed nodalization of the considered apparatus. In chemical reactors, however, the venting process is not only influenced by the flashing behaviour but additionally by the running chemical reaction in the vessel. Therefore the codes used for modelling have to consider the kinetics of the chemical reaction. Further multi-component systems and dissolving processes have to be regarded. In order to preduct the fluid- and thermodynamic process it could be helpful to use 3-dimensional codes in combination with the one-dimensional codes as used in nuclear industry to get a more detailed describtion of the running processes. (orig./HP)

  8. Pressure vessel for a BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

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

  9. Results of reactor pressure vessels ISI

    International Nuclear Information System (INIS)

    Cepcek, S.

    1994-01-01

    To find out the possible influence of the annealing process to reactor pressure vessel integrity, a large in-service inspection programme has been implemented as an associated activity to reactor pressure vessel annealing. In this paper the approach to the RPV in-service inspection is shown. Also, the main results and conclusions following in-service inspection are presented. (author). 3 refs, 1 fig

  10. Composite Overwrapped Pressure Vessels, A Primer

    Science.gov (United States)

    McLaughlan, Pat B.; Forth, Scott C.; Grimes-Ledesma, Lorie R.

    2011-01-01

    Due to the extensive amount of detailed information that has been published on composite overwrapped pressure vessels (COPVs), this document has been written to serve as a primer for those who desire an elementary knowledge of COPVs and the factors affecting composite safety. In this application, the word "composite" simply refers to a matrix of continuous fibers contained within a resin and wrapped over a pressure barrier to form a vessel for gas or liquid containment. COPVs are currently used at NASA to contain high pressure fluids in propulsion, science experiments, and life support applications. They have a significant weight advantage over all metal vessels but require unique design, manufacturing, and test requirements. COPVs also involve a much more complex mechanical understanding due to the interplay between the composite overwrap and the inner liner. A metallic liner is typically used in a COPV as a fluid permeation barrier. The liner design concepts and requirements have been borrowed from all-metal vessels. However, application of metallic vessel design standards to a very thin liner is not straightforward. Different failure modes exist for COPVs than for all-metal vessels, and understanding of these failure modes is at a much more rudimentary level than for metal vessels.

  11. Liquid Nitrogen Subcooler Pressure Vessel Engineering Note

    Energy Technology Data Exchange (ETDEWEB)

    Rucinski, R.; /Fermilab

    1997-04-24

    The normal operating pressure of this dewar is expected to be less than 15 psig. This vessel is open to atmospheric pressure thru a non-isolatable vent line. The backpressure in the vent line was calculated to be less than 1.5 psig at maximum anticipated flow rates.

  12. Analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

    Oliveira, C.A.N. de; Augusto, O.B.

    1985-01-01

    This work proposes a methodology for the structural analysis of high diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem, and the results are compared with results obtained by the finite element method. (Author) [pt

  13. Blood vessels, circulation and blood pressure.

    Science.gov (United States)

    Hendry, Charles; Farley, Alistair; McLafferty, Ella

    This article, which forms part of the life sciences series, describes the vessels of the body's blood and lymphatic circulatory systems. Blood pressure and its regulatory systems are examined. The causes and management of hypertension are also explored. It is important that nurses and other healthcare professionals understand the various mechanisms involved in the regulation of blood pressure to prevent high blood pressure or ameliorate its damaging consequences.

  14. Motor-driven screwing and transporting tool for pressure vessels, especially pressure vessels for nuclear reactors

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    A screwing and transporting device for tensioning and loosening the reactor pressure vessel head is described. The advantage of the tool is its ability to unscrew the stud bolts from the lower part of the pressure vessel, too, and therefore make them accessible for in-service inspection. (TK) [de

  15. Pressure vessel integrity and weld inspection procedure

    International Nuclear Information System (INIS)

    Solomon, K.A.; Okrent, D.; Kastenberg, W.E.

    1975-01-01

    The primary objective of this paper is to develop a simple methodology which, when coupled with existing observations on pressure vessel behavior, provides an inter-relation between pressure vessel integrity, and the parameters of the in-service inspection program, including inspection sample size, frequency and efficiency. A modified Markov process is employed and a computer code was written to obtain numerical results. The Markov process mathematically describes the following physical events. In a nuclear reactor pressure vessel weld, some defects may exist prior to the zeroth inspection (i.e., prior to vessel operation). During the zeroth inspection and repair processes, some of these defects are removed. During the first cycle of vessel operation, the existing defects may grow and some new defects may be generated. Those defects that are found at the first (and succeeding) inspection interval and warrant repair, are repaired. The above process continues through several operating cycles to the end of vessel life. During any inspection, only a portion of the welds may be inspected, and with less than perfect efficiency

  16. Design of Saddle Support for Horizontal Pressure Vessel

    OpenAIRE

    Vinod Kumar; Navin Kumar; Surjit Angra; Prince Sharma

    2014-01-01

    This paper presents the design analysis of saddle support of a horizontal pressure vessel. Since saddle have the vital role to support the pressure vessel and to maintain its stability, it should be designed in such a way that it can afford the vessel load and internal pressure of the vessel due to liquid contained in the vessel. A model of horizontal pressure vessel and saddle support is created in ANSYS. Stresses are calculated using mathematical approach and ANSYS soft...

  17. 46 CFR 50.30-15 - Class II pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class II pressure vessels. 50.30-15 Section 50.30-15... Fabrication Inspection § 50.30-15 Class II pressure vessels. (a) Class II pressure vessels shall be subject to... pressure vessels shall be performed during the welding of the longitudinal joint. At this time the marine...

  18. 46 CFR 50.30-20 - Class III pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Class III pressure vessels. 50.30-20 Section 50.30-20... Fabrication Inspection § 50.30-20 Class III pressure vessels. (a) Class III pressure vessels shall be subject... specifically exempted by other regulations in this subchapter. (b) For Class III welded pressure vessels, one...

  19. 46 CFR 61.10-5 - Pressure vessels in service.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessels in service. 61.10-5 Section 61.10-5... INSPECTIONS Tests and Inspections of Pressure Vessels § 61.10-5 Pressure vessels in service. (a) Basic requirements. Each pressure vessel must be examined or tested every 5 years. The extent of the test or...

  20. Curved and conformal high-pressure vessel

    Science.gov (United States)

    Croteau, Paul F.; Kuczek, Andrzej E.; Zhao, Wenping

    2016-10-25

    A high-pressure vessel is provided. The high-pressure vessel may comprise a first chamber defined at least partially by a first wall, and a second chamber defined at least partially by the first wall. The first chamber and the second chamber may form a curved contour of the high-pressure vessel. A modular tank assembly is also provided, and may comprise a first mid tube having a convex geometry. The first mid tube may be defined by a first inner wall, a curved wall extending from the first inner wall, and a second inner wall extending from the curved wall. The first inner wall may be disposed at an angle relative to the second inner wall. The first mid tube may further be defined by a short curved wall opposite the curved wall and extending from the second inner wall to the first inner wall.

  1. Empirical correlation between mechanical and physical parameters of irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Tipping, P.; Solt, G.; Waeber, W.

    1991-02-01

    Neutron irradiation embrittlement of nuclear reactor pressure vessel (PV) steels is one of the best known ageing factors of nuclear power plants. If the safety limits set by the regulators for the PV steel are not satisfied any more, and other measures are too expensive for the economics of the plant, this embrittlement could lead to the closure of the plant. Despite this, the fundamental mechanisms of neutron embrittlement are not yet fully understood, and usually only empirical mathematical models exist to asses neutron fluence effects on embrittlement, as given by the Charpy test for example. In this report, results of a systematic study of a French forging (1.2 MD 07 B), irradiated to several fluences will be reported. Mechanical property measurements (Charpy tensile and Vickers microhardness), and physical property measurements (small angle neutron scattering - SANS), have been done on specimens having the same irradiation or irradiation-annealing-reirradiation treatment histories. Empirical correlations have been established between the temperature shift and the decrease in the upper shelf energy as measured on Charpy specimens and tensile stresses and hardness increases on the one hand, and the size of the copper-rich precipitates formed by the irradiation on the other hand. The effect of copper (as an impurity element) in enhancing the degradation of mechanical properties has been demonstrated; the SANS measurements have shown that the size and amount of precipitates are important. The correlations represent the first step in an effort to develop a description of neutron irradiation induced embrittlement which is based on physical models. (author) 6 figs., 27 refs

  2. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1993-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described together with different aspects of the multielement integrity argument. The main revisions to the mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels are discussed. (author)

  3. Reactor Structural Materials: Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R

    2000-07-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported.

  4. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

  5. Integrity of Magnox reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Flewitt, P.E.J.; Williams, G.H.; Wright, M.B.

    1992-01-01

    The background to the safety assessment of the steel reactor pressure vessels for Magnox power stations is reviewed. The evolved philosophy adopted for the 1991 safety cases prepared for the continued operation of four Magnox power stations operated by Nuclear Electric plc is described, together with different aspects of the multi-legged integrity argument. The main revisions to the materials mechanical property data are addressed together with the assessment methodology adopted and their implications for the overall integrity argument formulated for the continued safe operation of these reactor pressure vessels. (author)

  6. Internal Friction of Pressure Vessel Steel Embrittlement

    International Nuclear Information System (INIS)

    Van Ouytsel, K.

    2001-01-01

    The contribution consists of an abstract of a PhD thesis. The thesis contains a literature study, a description of the construction details of a new inverted torsion pendulum. This device was designed to investigate pressure-vessel steels at high amplitudes (10 -4 to 10 -2 ) and over a wide temperature range (90-700K) at approximately 1 Hz in the irradiated condition. Results of measurements on a variety of reactor pressure vessel steels by means of the torsion penduli are reported and interpreted

  7. Proactive life extension of pressure vessels

    Science.gov (United States)

    Mager, Lloyd

    1998-03-01

    For a company to maintain its competitive edge in today's global market every opportunity to gain an advantage must be exploited. Many companies are strategically focusing on improved utilization of existing equipment as well as regulatory compliance. Abbott Laboratories is no exception. Pharmaceutical companies such as Abbott Laboratories realize that reliability and availability of their production equipment is critical to be successful and competitive. Abbott Laboratories, like many of our competitors, is working to improve safety, minimize downtime and maximize the productivity and efficiency of key production equipment such as the pressure vessels utilized in our processes. The correct strategy in obtaining these objectives is to perform meaningful inspection with prioritization based on hazard analysis and risk. The inspection data gathered in Abbott Laboratories pressure vessel program allows informed decisions leading to improved process control. The results of the program are reduced risks to the corporation and employees when operating pressure retaining equipment. Accurate and meaningful inspection methods become the cornerstone of a program allowing proper preventative maintenance actions to occur. Successful preventative/predictive maintenance programs must utilize meaningful nondestructive evaluation techniques and inspection methods. Nondestructive examination methods require accurate useful tools that allow rapid inspection for the entire pressure vessel. Results from the examination must allow the owner to prove compliance of all applicable regulatory laws and codes. At Abbott Laboratories the use of advanced NDE techniques, primarily B-scan ultrasonics, has provided us with the proper tools allowing us to obtain our objectives. Abbott Laboratories uses B-scan ultrasonics utilizing a pulse echo pitch catch technique to provide essential data on our pressure vessels. Equipment downtime is reduced because the nondestructive examination usually takes

  8. Summary of the NEA/CSNI Workshop on the safety assessment of reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Toerroenen, K. (VTT (Finland)); English, C.A. (AEA Tehnology (United Kingdom)); Crutzen, S. (CEC/TRC (Italy)); Hemsworth, B. (Health and Safety Executive (United Kingdom)); Noel, R.C. (EDF (France))

    1990-12-01

    The workshop was initiated by the CSNI Principal Working Group No 3, Reactor Component Integrity as part of its programme concerned with the long term degradation and service life management of reactor components. A central issue in this context and in consideration of reactor plant license extension is the capability to provide reliable safety assessments of component integrity using available data and assessment methods. In this regard, The Principal Working Group plans a systematic survey of reactor structural components critical to safety, commencing with pressure vessels. Existing knowledge and recent achievements in research on the behaviour of neutron irradiated materials, applications of fracture mechanics and probabilistic safety assessment methods provide comprehensive information to the regulatory bodies and plant operators. The Group considered that a workshop penetrating the problems met by authorities, utilities and researchers assessing the safety of reactor pressure vessels would clarify the current state of knowledge and the need and direction for further research. This report is a summary of the NEA/CSNI workshop. (au).

  9. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

    1983-01-01

    This construction of a container, which is pressure-relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 900 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW)

  10. Pressure vessel for gaseous media

    International Nuclear Information System (INIS)

    Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

    1977-01-01

    This construction of a container, which is pressure relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 90 0 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW) 891 RW [de

  11. Examination of VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Markulin, K.

    2008-01-01

    The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

  12. Reactor pressure vessel with forged nozzles

    Science.gov (United States)

    Desai, Dilip R.

    1993-01-01

    Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

  13. Procedures, methods and computer codes for the probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks

    International Nuclear Information System (INIS)

    Qian, Guian; Niffenegger, Markus

    2013-01-01

    The reactor pressure vessel (RPV), as one of the most important safety barriers of light water reactors, is exposed to neutron irradiation at elevated temperatures, which results in embrittlement of the RPV steel. One potential challenge to the structural integrity of the RPV in a pressurized water reactor is posed by pressurized thermal shock (PTS). Therefore, the safety of the RPV with regard to neutron embrittlement has to be analyzed. In this paper, the procedure and method for the structural integrity analysis of RPV subjected to PTS are presented. FAVOR and PASCAL, two computer codes widely used for the probabilistic analysis of RPV subjected to PTS, are briefly reviewed and compared. By using FAVOR, a benchmark example is presented to show the procedure and method for the integrity analysis. The influence of warm prestressing (WPS), fracture toughness and constraint effect on the integrity analysis of RPV is discussed. The Master Curve method is more realistic than the ASME model to consider the analysis of fracture toughness and thus is recommended. In order to transfer the fracture toughness data from test specimen to the RPV, local approach provides a probabilistic method

  14. Cylindrical pressure vessel constructed of several layers

    International Nuclear Information System (INIS)

    Yamauchi, Takeshi.

    1976-01-01

    For a cylindrical pressure vessel constructed of several layers whose jacket has at least one circumferential weld joining the individual layers, it is proposed to provide this at least at the first bending line turning point (counting from the weld between the jacket and vessel floor), which the sinusoidally shaped jacket has. The section of the jacket extending in between should be made as a full wall section. The proposal is based on calculations of the bending stiffness of cylindrical jackets, which could not yet be confirmed for jackets having several layers. (UWI) [de

  15. Revisiting the reactor pressure vessel for long-time operation; Revisitando la vasija a presion del reactor para largos tiempos de operacion

    Energy Technology Data Exchange (ETDEWEB)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-07-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIFFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  16. Heat treatment device for extending the life of a pressure vessel, particularly a reactor pressure vessel

    International Nuclear Information System (INIS)

    Krauss, P.; Mueller, E.; Poerner, H.; Weber, R.

    1979-01-01

    A support body in the form of an insulating cylinder is tightly sealed by connected surfaces at its outer circumference to the inner wall of the pressure vessel. It forms an annular heating space. The heat treatment or tempering of the pressure vessel takes place with the reactor space empty and screened from the outside by ceiling bolts. Heating gas or an induction winding can be used as the means of heating. (DG) [de

  17. A framework expert system for pressure vessels

    International Nuclear Information System (INIS)

    Wang, Y.C.; Qin, S.J.

    1989-01-01

    Expert systems, known as a powerful tool to those numerical problems accompanied with logical argumentation, are facing the era of extended application into the engineering fields beyond the classical scopes of diagnosis and consultation. With regard to pressure vessels design it seems that the most important task is to establish a general purpose frame based on a microcomputer skeleton system to meet the various requirements of different vessels. The authors have made an attempt to perform such a skeleton designated file, ESTOOL, in order to achieve the objectives of executing numerical calculation combined with logical reasoning, and attaining higher efficiency of rules searching process. It has been successfully patched to the design software package for jacketed vessel with stirring shaft. This paper presents the guiding concepts and basic structure of ESTOOL via knowledge acquisition subsystem and inference engine

  18. Thermal annealing in neutron-irradiated tribromobenzenes

    DEFF Research Database (Denmark)

    Siekierska, K.E.; Halpern, A.; Maddock, A. G.

    1968-01-01

    The distribution of 82Br among various products in neutron-irradiated isomers of tribromobenzene has been investigated, and the effect of thermal annealing examined. Reversed-phase partition chromatography was employed for the determination of radioactive organic products, and atomic bromine...

  19. Conformable pressure vessel for high pressure gas storage

    Science.gov (United States)

    Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

    2016-01-12

    A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

  20. Irradiation embrittlement of pressure vessel steels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Vacek, M.

    1975-01-01

    A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

  1. Composite Pressure Vessel Including Crack Arresting Barrier

    Science.gov (United States)

    DeLay, Thomas K. (Inventor)

    2013-01-01

    A pressure vessel includes a ported fitting having an annular flange formed on an end thereof and a tank that envelopes the annular flange. A crack arresting barrier is bonded to and forming a lining of the tank within the outer surface thereof. The crack arresting barrier includes a cured resin having a post-curing ductility rating of at least approximately 60% through the cured resin, and further includes randomly-oriented fibers positioned in and throughout the cured resin.

  2. Thermal ratcheting in pressure vessels and piping

    International Nuclear Information System (INIS)

    Kalnins, A.; Updike, D.P.

    1975-01-01

    During its lifetime, a nuclear power plant can experience cyclic thermal loading produced during start-ups and shut-downs. Under such conditions, pressure vessels, piping and other components can experience accumulative dimensional changes through thermal ratcheting. Such dimensional changes can occur when cyclic thermal loading is superposed upon steady mechanical loading. Simplified models based on one-dimensional plastic stress-strain relations, as proposed by Miller, Bree and Burgreen, are currently used by designers to estimate the likelihood and the amount of ratcheting during the life of the component. It is shown in this paper that the Miller-Bree-Burgreen one-dimensional model is applicable when the ratio of the steady membrane stresses is larger than one-half (Nsub(y)/Nsub(x)>1/2), which includes such significant applications as cylindrical and spherical vessels under internal pressure. For the stress ratios approaching minus one, the biaxial model predicts much higher ratchet strain. Examples where such states occur are the knuckle region in pressure vessels with torispherical heads and pipe connections under torsion. Designers should be alerted to such a limitation of the presently used ratcheting model. (Auth.)

  3. Code boiler and pressure vessel life assessment

    International Nuclear Information System (INIS)

    Farr, J.R.

    1992-01-01

    In the United States of America and in Canada, laws and controls for determining life assessment for continued operation of equipment exist only for those pressure vessels built to Section III and evaluated according to Section XI. In this presentation, some of those considerations which are made in the USA and Canada for deciding on life or condition assessment of boilers and pressure vessels designed and constructed to other sections of the ASME Boiler and Pressure Vessel Code are reviewed. Life assessment or condition assesssment is essential in determining what is necessary for continued operation. With no ASME rules being adopted by laws or regulations, other than OSHA in the USA and similar environmental controls in Canada, to control life assessment for continued operation, the equipment owner must decide if assessment is to be done and how much to do. Some of those considerations are reviewed along with methods and procedures to make an assessment along with a discussion of where the ASME B and PV Code currently stands regarding continued operation. (orig.)

  4. Compact insert design for cryogenic pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, Salvador M.; Ledesma-Orozco, Elias Rigoberto; Espinosa-Loza, Francisco; Petitpas, Guillaume; Switzer, Vernon A.

    2017-06-14

    A pressure vessel apparatus for cryogenic capable storage of hydrogen or other cryogenic gases at high pressure includes an insert with a parallel inlet duct, a perpendicular inlet duct connected to the parallel inlet. The perpendicular inlet duct and the parallel inlet duct connect the interior cavity with the external components. The insert also includes a parallel outlet duct and a perpendicular outlet duct connected to the parallel outlet duct. The perpendicular outlet duct and the parallel outlet duct connect the interior cavity with the external components.

  5. Innovations in prestressed concrete pressure vessel design

    International Nuclear Information System (INIS)

    Chow, P.Y.; Ngo, D.; Lin, T.Y.

    1979-01-01

    The study explored a new approach to the design of a high-pressure PCPV that accepts tension and tension cracks in the outer region of the PCPV. It examined the possibility of incorporating artificially-introduced preformed separations that pre-determined crack locations in the design as a method of controlling high tensile stresses generated by internal temperature and pressure. The results showed that the PCPV so designed was, in the extreme case of the DSV, approximately 70% cheaper than the 18 steel vessels of equivalent capacity it replaces. (orig.)

  6. Safety of steel vessel Magnox pressure circuits

    International Nuclear Information System (INIS)

    Stokoe, T.Y.; Bolton, C.J.; Heffer, P.J.H.

    1991-01-01

    The maintenance of pressure circuit integrity is fundamental to nuclear safety at the steel vessel Magnox stations. To confirm continued pressure circuit integrity the CEGB, as part of the Long Term Safety Review, has carried out extensive assessment and inspection in recent years. The assessment methods and inspection techniques employed are based on the most modern available. Reactor pressure vessel integrity is confirmed by a combination of arguments including safety factors inferred from the successful pre-service overpressure test, leak-before-break analysis and probabilistic assessment. In the case of other parts of the pressure circuits that are more accessible, comprising the boiler shells and interconnecting gas duct work, in-service inspection is a major element of the safety substantiation. The assessment and inspection techniques and the materials property data have been underpinned for many years by extensive research and development programmes and in-reactor monitoring of representative samples has also been undertaken. The paper summarises the work carried out to demonstrate the long term integrity of the Magnox pressure circuits and provides examples of the results obtained. (author)

  7. A study of the pressure vessel steel of the WWER-440 unit 1 of the Kozloduy nuclear power plant

    Science.gov (United States)

    Kostadinova, E.; Velinov, N.; Avdjieva, T.; Mitov, I.; Rusanov, V.

    2017-11-01

    A comparison between highly neutron irradiated samples from the region of weld № 4 and low irradiated samples from weld № 1 taken from the pressure vessel of the WWER-440 Unit № 1 of the Kozloduy NPP has been performed. Measurements of the residual activity of samples from the outer surface of the reactor pressure vessel bottom corpus reveal very low activity of 60Co. Insofar as there the base and weld metal appear to be exposed to a very low neutron fluence, the samples from these locations can be considered as practically not affected and may serve as a reference basis for comparison with highly irradiated pressure vessel regions. The Mössbauer parameters isomer shift (IS) and quadrupole splitting (QS) were found to be absolutely irradiation insensitive. A stepwise reduction of the internal hyperfine magnetic field Bhf, each by about 2.6 T, was observed. This can be attributed to the replacement of one or two surrounding iron atoms as first nearest neighbors by non-iron alloying atoms. The Mössbauer experimental line widths for irradiated and non-irradiated samples are practically the same, which is a quite unexpected result. The area fraction ratio for the three main Zeeman sextet subspectra S1:S2:S3 shows very high irradiation sensitivity. For the bottom low irradiated region of the reactor vessel the values are S1:S2:S3 = 50.1:40.0:9.4. After seven years of operation between the pressure vessel annealing in 1989 and the autumn of 1996 when the samples from weld № 4 were taken the ratio changes strongly to S1:S2:S3 = 56.4:34.7:8.5. A possible explanation of this result is that neutron irradiation gives rise to a precipitation process involving predominantly alloying atoms as Ni, Mn, Cr, Mo and V which become mobile and precipitate in the form of carbides and/or P-rich phases and alloying atom aggregates. This "refinement" process lowers the partial area of subspectra S2 and S3 where alloying atoms are involved and leads to a higher area fraction of

  8. Modeling copper precipitation hardening and embrittlement in a dilute Fe-0.3at.%Cu alloy under neutron irradiation

    Science.gov (United States)

    Bai, Xian-Ming; Ke, Huibin; Zhang, Yongfeng; Spencer, Benjamin W.

    2017-11-01

    Neutron irradiation in light water reactors can induce precipitation of nanometer sized Cu clusters in reactor pressure vessel steels. The Cu precipitates impede dislocation gliding, leading to an increase in yield strength (hardening) and an upward shift of ductile-to-brittle transition temperature (embrittlement). In this work, cluster dynamics modeling is used to model the entire Cu precipitation process (nucleation, growth, and coarsening) in a Fe-0.3at.%Cu alloy under neutron irradiation at 300°C based on the homogenous nucleation mechanism. The evolution of the Cu cluster number density and mean radius predicted by the modeling agrees well with experimental data reported in literature for the same alloy under the same irradiation conditions. The predicted precipitation kinetics is used as input for a dispersed barrier hardening model to correlate the microstructural evolution with the radiation hardening and embrittlement in this alloy. The predicted radiation hardening agrees well with the mechanical test results in the literature. Limitations of the model and areas for future improvement are also discussed in this work.

  9. Dislocation-defect interactions in nuclear reactor pressure-vessel steels investigated by means of internal friction

    Energy Technology Data Exchange (ETDEWEB)

    Van Ouytsel, K. [CEN, Mol (Belgium). Nucl. Res. Centre SCK; De Batist, R. [University of Antwerp RUCA, Middelheimlaan 1, 2020, Antwerp (Belgium); Schaller, R. [Institut de Genie Atomique, Ecole Polytechnique Federale de Lausanne, PHB Ecublens, CH-1015, Lausanne (Switzerland)

    2000-09-28

    A study of pressure-vessel steel embrittlement mechanisms by means of temperature-dependent and amplitude-dependent internal friction has been carried out within the framework of commercial surveillance of nuclear reactor components. An inverted torsion pendulum operating at {proportional_to}1 Hz has been employed to study a wide variety of pressure-vessel steels and an IAEA reference material in various conditions. This contribution will discuss the results for the JRQ reference material only and serve as a basis on which to interpret the data from real pressure-vessel steels. The temperature-dependent experiments evidence a reduction in the dislocation mobility as a result of neutron irradiation and prove that the technique is sensitive to thermal ageing involving changes in the dislocation mobility and type of dislocation-defect interaction. Amplitude-dependent internal friction provides a means to determine the yield strength of the material. The importance of the influence of dislocation dragging on the yield stress is highlighted. (orig.)

  10. 46 CFR 115.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Pressure vessels and boilers. 115.812 Section 115.812... CERTIFICATION Material Inspections § 115.812 Pressure vessels and boilers. (a) Pressure vessels must be tested... testing requirements for boilers are contained in § 61.05 in subchapter F of this chapter. [CGD 85-080, 61...

  11. 46 CFR 58.60-3 - Pressure vessel.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessel. 58.60-3 Section 58.60-3 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) MARINE ENGINEERING MAIN AND AUXILIARY MACHINERY AND... Pressure vessel. A pressure vessel that is a component in an industrial system under this subpart must meet...

  12. Leak detector for reactor pressure vessel

    International Nuclear Information System (INIS)

    Morimoto, Mikio.

    1991-01-01

    A branched pipe is disposed to a leak off pipeline led from a flange surface which connects the main body and the upper lid of a reactor pressure vessel. An exhaust pump is disposed to the branched pipe and a moisture gage is disposed on the side of the exhaustion and a dry air supplier is connected to the branched pipe. Upon conducting a pressure-proof leak test for the reactor pressure vessel, the exhaust pump is operated and an electromagnet valve disposed at the upstream of the dry air supplier is opened and closed repeatedly. The humidity of air sucked by the exhaust pump is detected by the moisture gage. If leaks should be caused in the joining surface of the flange, leaked water is diffused as steams. Accordingly, occurrence of leak can be detected instantly based on the comparison with the moisture level of the dry air as a standard. In this way, a leak test can be conducted reliably in a short period of time with no change of for the reactor pressure container itself. (I.N.)

  13. Nuclear reactor pressure vessel support system

    Science.gov (United States)

    Sepelak, George R.

    1978-01-01

    A support system for nuclear reactor pressure vessels which can withstand all possible combinations of stresses caused by a postulated core disrupting accident during reactor operation. The nuclear reactor pressure vessel is provided with a flange around the upper periphery thereof, and the flange includes an annular vertical extension formed integral therewith. A support ring is positioned atop of the support ledge and the flange vertical extension, and is bolted to both members. The plug riser is secured to the flange vertical extension and to the top of a radially outwardly extension of the rotatable plug. This system eliminates one joint through which fluids contained in the vessel could escape by making the fluid flow path through the joint between the flange and the support ring follow the same path through which fluid could escape through the plug risers. In this manner, the sealing means to prohibit the escape of contained fluids through the plug risers can also prohibit the escape of contained fluid through the securing joint.

  14. Computing the partial volume of pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Wiencke, Bent [Nestle USA, Corporate Engineering, 800 N. Brand Blvd, Glendale, CA 91203 (United States)

    2010-06-15

    The computation of the partial and total volume of pressure vessels with various type of head profiles requires detailed knowledge of the head profile geometry. Depending on the type of head profile the derivation of the equations can become very complex and the calculation process cumbersome. Certain head profiles require numerical methods to obtain the partial volume, which for most application is beyond the scope of practicability. This paper suggests a unique method that simplifies the calculation procedure for the various types of head profiles by using one common set of equations without the need for numerical or complex computation methods. For ease of use, all equations presented in this paper are summarized in a single table format for horizontal and vertical vessels. (author)

  15. Recent experience reactor pressure vessel manufacture

    International Nuclear Information System (INIS)

    Vignes, A.

    1985-01-01

    This paper present the Framatome's recent experience in the manufacture of 1300 MWe PWR vessels; one shows how the very high standards of quality have been obtained to meet the stringent requirements. After a description of a pressure vessel, materials and forgings properties are presented. The nature and sequence of the main fabrication operations are reviewed. This paper deals after with the quality of welds, the preheating and post-heating equipment, the submerged arc welding process and procedures, the cladding process, and the under-clad cracking problems. Ultrasonic inspection procedures of the main welds are described with a comparison of RCCM (design and construction rules for mechanical components of PWR units) and Sizewell B specifications. Support of data on the reproductibility and effectiveness of ultrasonic examination and on the reliability given by repetitive inspection are presented

  16. Support device in a pressure vessel, particularly a reactor pressure vessel, to support against horizontal forces

    International Nuclear Information System (INIS)

    Dorner, H.; Harand, E.; Scholz, M.; Scheler, R.

    1986-01-01

    In a support device on a reactor pressure vessel of a PWR, a guide leading downwards is provided in the vertical pellet cartridge on the floor side of the reactor pressure vessel, which is located in a fixed support pipe. The support pipe is supported on a concreted baseplate in the horizontal direction, and can be screwed to this. The support pipe has a plate-shaped support foot to support it on the baseplate. The reactor pressure vessel, normally resting on its main coolant duct in the plane normal to the axis, can be effectively supported against horizontal forces by this additional support, without preventing thermal movement in the longitudinal axis. (orig./HP) [de

  17. Dictionary of pressure vessel and piping technology

    International Nuclear Information System (INIS)

    Jentgen, L.; Schmitz, H.P.

    1986-01-01

    A specialised dictionary has been compiled containing the appropriate English and German terms in the following technical fields: materials science, welding, destructive and non-destructive testing, thermal and mass transfer, the design and construction in particular of pressure vessels, tanks, heat exchangers, piping, expansion joints, valves, and components associated with the above fields. This dictionary is the result of many years spent in evaluating technical terminology from the relevant American and British regulations, technical rules, standards, and specifications (see bibliography) and correlating these with the terminology of comparable German regulations, rules and standards, together with the essential technical literature. (orig.) [de

  18. Reactor pressure vessel aging and countermeasures

    International Nuclear Information System (INIS)

    Leitz, C.

    1987-01-01

    The considerable aging effect on reactor pressure vessels is the effect of irradiation on material properties in the core beltline region. Modern LWRs in the Federal Republic of Germany are designed such that irradiation effects are very low. Countermeasures applicable separately or in combination for plants with higher than normally expected irradiation effects are described in three steps: first, examinations and calculations to extend the formal reactor lifetime by reducing over-conservative margins; second, changes in core design to reduce future irradiation effects; third, a procedure to recover irradiation effect on material properties already sustained. (orig./HP) [de

  19. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  20. Reactor Pressure Vessel (RPV) Acquisition Strategy

    Energy Technology Data Exchange (ETDEWEB)

    Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2008-04-01

    The Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Reactor Pressure Vessel (RPV). This component will be larger than any nuclear reactor pressure vessel presently in service in the United States. The RPV will be taller, larger in diameter, thicker walled, heavier and most likely fabricated at the Idaho National Laboratory (INL) site of multiple subcomponent pieces. The pressure vessel steel can either be a conventional materials already used in the nuclear industry such as listed within ASME A508/A533 specifications or it will be fabricated from newer pressure vessel materials never before used for a nuclear reactor in the US. Each of these characteristics will present a

  1. An internal-friction study of reactor-pressure-vessel steel embrittlement

    International Nuclear Information System (INIS)

    Ouytsel, K. van; Fabry, A.; Batist, R. de; Schaller, R.

    1997-01-01

    Within an enhanced commercial surveillance strategy, the nuclear-research institute SCK.CEN in Mol, Belgium is investigating, by means of internal friction, the microstructural processes responsible for embrittlement of pressure-vessel steels. The experiments were carried out using a torsion pendulum at the Ecole Polytechnique Federale de Lausanne in Switzerland. Amplitude-independent internal-friction experiments teach us that neutron irradiation induces defects which interact with mobile dislocations. Thermal ageing of JRQ and Doel-IV steel does not cause major embrittlement effects. Amplitude-dependent internal-friction experiments allow us to determine a critical amplitude which corresponds to the yield stress of the material as obtained from static tensile tests. The results also correspond to a three-component model for the yield strength taking into account both hardening and non-hardening embrittlement. Investigations of Doel-I-II weld material in different conditions reveal that embrittlement due to irradiation or thermal ageing can be interpreted in terms of a fine interplay between long- and short-range phenomena. (author)

  2. Life prediction study of reactor pressure vessel as essential technical foundation for plant life extension

    International Nuclear Information System (INIS)

    Nakajima, H.; Nakajima, N.; Kondo, T.

    1987-01-01

    The life of a LWR plant is determined essentially by the limit of reliable performance of the components which are difficult to replace without high economic and/or safety risks. Typical of such a component is the reactor pressure vessel (RPV). The engineering life of a RPV of a given quality of steel is considered to be a complex function of factors such as the resistance to fracture, which has deteriorated due to neutron irradiation and thermal aging, and generation of surface flaws by environmental effects such as corrosion and their growth under operational load that varies during steady state operation and transients. In an attempt to evaluate the engineering life of a RPV of a LWR, a preliminary survey was made by applying a set of knowledge accumulated primarily in the field of subcritical crack growth behavior of RPV steels in reactor water environments. The major conclusions drawn are: (1) the life of a RPV is dependent on the quality of steel used, particularly with respect to any minor impurities it contains. (2) The issue of plant life extension in RPV aspect is found to be optimistic for cases where the steels used satisfy a reasonable level of quality control. (3) The importance of providing sound scientific foundation is stressed for the implementation of a practicable life extension scheme: this can be established through intensified studies of flaw growth and fracture behaviours in well defined testings under reasonably simulated service conditions

  3. 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

  4. Inservice inspection of Halden BWR pressure vessel

    International Nuclear Information System (INIS)

    Foerli, O.; Hernes, T.

    1978-01-01

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

  5. Neural Network Burst Pressure Prediction in Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Hill, Eric v. K.; Dion, Seth-Andrew T.; Karl, Justin O.; Spivey, Nicholas S.; Walker, James L., II

    2007-01-01

    Acoustic emission data were collected during the hydroburst testing of eleven 15 inch diameter filament wound composite overwrapped pressure vessels. A neural network burst pressure prediction was generated from the resulting AE amplitude data. The bottles shared commonality of graphite fiber, epoxy resin, and cure time. Individual bottles varied by cure mode (rotisserie versus static oven curing), types of inflicted damage, temperature of the pressurant, and pressurization scheme. Three categorical variables were selected to represent undamaged bottles, impact damaged bottles, and bottles with lacerated hoop fibers. This categorization along with the removal of the AE data from the disbonding noise between the aluminum liner and the composite overwrap allowed the prediction of burst pressures in all three sets of bottles using a single backpropagation neural network. Here the worst case error was 3.38 percent.

  6. Seal analysis technology for reactor pressure vessel

    International Nuclear Information System (INIS)

    Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

    2009-01-01

    There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)

  7. Pool critical assembly pressure vessel facility benchmark

    International Nuclear Information System (INIS)

    Remec, I.; Kam, F.B.K.

    1997-07-01

    This pool critical assembly (PCA) pressure vessel wall facility benchmark (PCA benchmark) is described and analyzed in this report. Analysis of the PCA benchmark can be used for partial fulfillment of the requirements for the qualification of the methodology for pressure vessel neutron fluence calculations, as required by the US Nuclear Regulatory Commission regulatory guide DG-1053. Section 1 of this report describes the PCA benchmark and provides all data necessary for the benchmark analysis. The measured quantities, to be compared with the calculated values, are the equivalent fission fluxes. In Section 2 the analysis of the PCA benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed for three ENDF/B-VI-based multigroup libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent agreement of the calculated (C) and measures (M) equivalent fission fluxes was obtained. The arithmetic average C/M for all the dosimeters (total of 31) was 0.93 ± 0.03 and 0.92 ± 0.03 for the SAILOR-95 and BUGLE-96 libraries, respectively. The average C/M ratio, obtained with the BUGLE-93 library, for the 28 measurements was 0.93 ± 0.03 (the neptunium measurements in the water and air regions were overpredicted and excluded from the average). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used

  8. Feedwater control device for reactor pressure vessels

    International Nuclear Information System (INIS)

    Oonuma, Takeshi.

    1982-01-01

    Purpose: To prevent the generation of thermal stresses at the junction between a clean-up water pipe and a feedwater pipe. Constitution: Hot water containing impurities in a pressure vessel is caused to flow by a recycling pump through a heat exchanger, a cooler and a clean-up desalter and again by way of the heat exchanger into the feedwater pipe at the junction with the clean-up water pipe, where it is mixed with the feedwater passed by way of a feedwater heater and supplied to the pressure vessel. The feedwater temperature for the feedwater pipe and the set temperature for the clean-up water are compared with each other by using temperature sensors disposed to the feedwater pipe between the junction and the feedwater heater at the upstream of the junction. If the temperature difference is increased, for instance, upon transient state where the operation of the feedwater heater is not yet stabilized, the recycling pump is controlled to stop the supply of the clean-up water to the junction while flowing only the feedwater. This makes the temperature distribution uniform and prevents the generation of the thermal stresses at the junction, by which reactor safety can be improved. (Moriyama, K.)

  9. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  10. Pressurized wet digestion in open vessels (T11)

    International Nuclear Information System (INIS)

    Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

    2002-01-01

    Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

  11. Chemical Safety Alert: Rupture Hazard of Pressure Vessels

    Science.gov (United States)

    Pressure vessels or boilers can fail catastrophically if they are not properly designed, constructed, operated, inspected, tested, or repaired. Risk increases if vessels contents are toxic, corrosive, reactive, or flammable.

  12. Reactor pressure vessel stud management automation strategies

    International Nuclear Information System (INIS)

    Biach, W.L.; Hill, R.; Hung, K.

    1992-01-01

    The adoption of hydraulic tensioner technology as the standard for bolting and unbolting the reactor pressure vessel (RPV) head 35 yr ago represented an incredible commitment to new technology, but the existing technology was so primitive as to be clearly unacceptable. Today, a variety of approaches for improvement make the decision more difficult. Automation in existing installations must meet complex physical, logistic, and financial parameters while addressing the demands of reduced exposure, reduced critical path, and extended plant life. There are two generic approaches to providing automated RPV stud engagement and disengagement: the multiple stud tensioner and automated individual tools. A variation of the latter would include the handling system. Each has its benefits and liabilities

  13. Nanostructure evolution under irradiation of Fe(C)MnNi model alloys for reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chiapetto, M., E-mail: mchiapet@sckcen.be [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium); Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Becquart, C.S. [Unité Matériaux Et Transformations (UMET), UMR 8207, Université de Lille 1, ENSCL, F-59600 Villeneuve d’Ascq Cedex (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Domain, C. [EDF R& D, Département Matériaux et Mécanique des Composants, Les Renardières, F-77250 Moret sur Loing (France); Laboratoire commun EDF-CNRS Etude et Modélisation des Microstructures pour le Vieillissement des Matériaux (EM2VM) (France); Malerba, L. [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, B-2400 Mol (Belgium)

    2015-06-01

    Radiation-induced embrittlement of bainitic steels is one of the most important lifetime limiting factors of existing nuclear light water reactor pressure vessels. The primary mechanism of embrittlement is the obstruction of dislocation motion produced by nanometric defect structures that develop in the bulk of the material due to irradiation. The development of models that describe, based on physical mechanisms, the nanostructural changes in these types of materials due to neutron irradiation are expected to help to better understand which features are mainly responsible for embrittlement. The chemical elements that are thought to influence most the response under irradiation of low-Cu RPV steels, especially at high fluence, are Ni and Mn, hence there is an interest in modelling the nanostructure evolution in irradiated FeMnNi alloys. As a first step in this direction, we developed sets of parameters for object kinetic Monte Carlo (OKMC) simulations that allow this to be done, under simplifying assumptions, using a “grey alloy” approach that extends the already existing OKMC model for neutron irradiated Fe–C binary alloys [1]. Our model proved to be able to describe the trend in the buildup of irradiation defect populations at the operational temperature of LWR (∼300 °C), in terms of both density and size distribution of the defect cluster populations, in FeMnNi model alloys as compared to Fe–C. In particular, the reduction of the mobility of point-defect clusters as a consequence of the presence of solutes proves to be key to explain the experimentally observed disappearance of detectable point-defect clusters with increasing solute content.

  14. Reinforced-concrete pressure vessel for a nuclear power plant

    International Nuclear Information System (INIS)

    Schwiers, H.G.; Schoening, J.

    1979-01-01

    The gas turbo-engine of the THTR-300 is installed in a horizontal duct of the prestressed concrete pressure vessel. The cavern and its recesses for the steam generators are arranged above and laterally away from this duct. By this means prestressing of the individual regions of the pressure vessel may be adapted to the pressure existing in the different cavities. (RW) [de

  15. Effect of neutron irradiation on response of reinforced concrete members for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, Kyoungsoo; Kim, Hyung-Tae [Department of Civil and Environmental Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 120-749 (Korea, Republic of); Kwon, Tae-Hyun, E-mail: taekwonkr@kaeri.re.kr [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Choi, Eunsoo [Department of Civil and Environmental Engineering, Hongik University, 94 Wausan-ro, Mapo-gu, Seoul 121-791 (Korea, Republic of)

    2016-12-15

    Highlights: • Effects of long-term irradiation on reinforced concrete (RC) structures were investigated. • Responses of irradiated RC members were numerically investigated in terms of ductility. • Results demonstrated that energy dissipation capacity decreased under radiation environment. • Level of neutron radiation could be critical for RC structures during operation. - Abstract: In this study, the effects of long-term irradiation on the behaviors of reinforced concrete (RC) members were investigated to obtain a better understanding of the behaviors of RC structures under an irradiation environment, which include the biological shield walls and reactor vessel support structures of nuclear power plants (NPPs). The behaviors of three RC members were examined (a beam, beam-column section, and column under cyclic loading) by considering the changes in the constituent material properties due to neutron irradiation. The load capacity generally increases for a tension failure member with an increase in neutron irradiation because neutron irradiation increases the yield stress of reinforcing steel. However, the load capacity of a compression failure member decreases with a decrease in the compressive strength of concrete when the fluence of neutron radiation increases. Additionally, RC member analysis results demonstrate that the energy dissipation capacity, which is a critical factor in seismic design, decreases significantly when the fluence of neutron radiation is greater than 1.0 × 10{sup 17} n/cm{sup 2}. Therefore, the level of neutron irradiation could be critical for RC structures over the long-term operation of NPPs, and thus the effects of neutron irradiation on RC structures should be considered as age-related damage.

  16. Exploratory Study of Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chernobaeva, A.A., Kryukov, A.M., Nikolaev, Y.A., Korolev, Y.N. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation)], Sokolov, M.A., Nanstad, R.K. [Oak Ridge National Lab., TN (United States)

    1997-12-31

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVS) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The working group agreed that each side would irradiate, anneal, reirradiate (if feasible), and test two materials of the other; so far, only charpy impact and tensile specimens have been included. Oak Ridge National Laboratory (ornl) conducted such a program (irradiation and annealing) with two weld metals representative of VVER-440 AND VVER-1000 RPVS, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation,annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) program plate 02 and Heavy-Section Steel Irradiation (HSSI) program weld 73w. The results for each material from each laboratory are compared with those from the other laboratory. the ORNL experiments with the VVER welds included irradiation to about 1 x 10 (exp 19) N/SQ CM ({gt}1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 X 10 (exp 19) N/SQ CM ({gt}1 MeV).

  17. Investigation of hydrogen isotopes interaction processes with lithium under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Zaurbekova, Zhanna, E-mail: zaurbekova@nnc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Skakov, Mazhyn; Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Tazhibayeva, Irina; Baklanov, Viktor; Barsukov, Nikolay [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Chikhray, Yevgen [Institute of Experimental and Theoretical Physics of Kazakh National University, Almaty (Kazakhstan)

    2016-11-01

    Highlights: • The experiments on study of helium and tritium generation and release processes under neutron irradiation from lithium saturated with deuterium are described in paper. ​ • The values of relative tritium and helium yield from lithium sample at different levels of neutron irradiation is calculated. • It was concluded that the main affecting process on tritium release from lithium is its interaction with lithium atoms with formation of lithium tritide. - Abstract: The paper describes the experiments on study of helium and tritium generation and release processes from lithium saturated with deuterium under neutron irradiation (in temperature range from 473 to 773 K). The diagrams of two reactor experiments show the time dependences of helium, DT, T{sub 2}, and tritium water partial pressures changes in experimental chamber with investigated lithium sample. According to experimental results, the values of relative tritium and helium yield from lithium sample at different levels of neutron irradiation were calculated. The time dependences of relative tritium and helium yield from lithium sample were plotted. It was concluded that the main affecting process on tritium release from lithium is its interaction with lithium atoms with formation of lithium tritide.

  18. Optical absorption characteristics of neutron irradiated heavy metal fluoride glasses

    Energy Technology Data Exchange (ETDEWEB)

    Mitra, S.S.; Banerjee, P.K.; Pereira, J.M.T.; Gedam, S.G.

    1987-10-15

    Samples of ZBLA and HBLA glasses were subjected to various fluences of neutron irradiation, and the spectral dependence of optical absorption was measured before and after irradiation. The IR edge was found to be unaffected by neutron irradiation for the fluences used. However, a red shift occurred at the UV edge which slightly recovered after three weeks.

  19. Impact of neutron irradiation on thermal helium desorption from iron

    Science.gov (United States)

    Hu, Xunxiang; Field, Kevin G.; Taller, Stephen; Katoh, Yutai; Wirth, Brian D.

    2017-06-01

    The synergistic effect of neutron irradiation and transmutant helium production is an important concern for the application of iron-based alloys as structural materials in fission and fusion reactors. In this study, we investigated the impact of neutron irradiation on thermal helium desorption behavior in high purity iron. Single crystalline and polycrystalline iron samples were neutron irradiated in HFIR to 5 dpa at 300 °C and in BOR-60 to 16.6 dpa at 386 °C, respectively. Following neutron irradiation, 10 keV He ion implantation was performed at room temperature on both samples to a fluence of 7 × 1018 He/m2. Thermal desorption spectrometry (TDS) was conducted to assess the helium diffusion and clustering kinetics by analyzing the desorption spectra. The comparison of He desorption spectra between unirradiated and neutron irradiated samples showed that the major He desorption peaks shift to higher temperatures for the neutron-irradiated iron samples, implying that strong trapping sites for He were produced during neutron irradiation, which appeared to be nm-sized cavities through TEM examination. The underlying mechanisms controlling the helium trapping and desorption behavior were deduced by assessing changes in the microstructure, as characterized by TEM, of the neutron irradiated samples before and after TDS measurements.

  20. Flux dependence of cluster formation in neutron-irradiated weld material

    International Nuclear Information System (INIS)

    Bergner, F; Ulbricht, A; Hein, H; Kammel, M

    2008-01-01

    The effect of neutron flux on the formation of irradiation-induced clusters in reactor pressure vessel (RPV) steels is an unresolved issue. Small-angle neutron scattering was measured for a neutron-irradiated RPV weld material containing 0.22 wt% impurity Cu. The experiment was focused on the influence of neutron flux on the formation of irradiation-induced clusters at fixed fluence. The aim was to separate and tentatively interpret the effect of flux on the characteristics of the cluster size distribution. We have observed a pronounced effect of neutron flux on cluster size, whereas the total volume fraction of irradiation-induced clusters is insensitive to the level of flux. The result is compatible with a rate theory model according to which the range of applied fluxes covers the transition from a flux-independent regime at lower fluxes to a regime of decelerating cluster growth. The results are confronted with measured irradiation-induced changes of mechanical properties. Despite the observed flux effect on cluster size, both yield stress increase and transition temperature shift turned out to be independent of flux. This is in agreement with the volume fraction of irradiation-induced clusters being insensitive to the level of flux

  1. Current Amplification Characteristics of BJT on Fast Neutron Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sung Ho; Sun, Gwang Min; Baek, Hani [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    BJT (Bipolar Junction Transistor) is a three-terminal device with an important feature in that the current through two terminals can be controlled by small changes we make in the current or voltage at the third terminal. This control feature allows us to amplify small AC signals or to switch the device from an on state and off state and back. Fast neutron irradiation incurs lattice damage in bulk Si. The recombination rate of minority carriers and register are increased by the lattice damage. This study will investigate the current amplification characteristics of a pnp Si BJT through fast neutron irradiation experiments. In this paper, the current amplification characteristics of a pnp Si BJT were investigated for fast neutron irradiation. The experimental results show that base-tocollector current amplification ratio is decreased with an increase in the fast neutron irradiation. These indicate that the lattice damage caused by fast neutron irradiation increases the recombination rate of minority carriers and resistor.

  2. Plastic collapse pressure of cylindrical vessels containing longitudinal surface cracks

    Energy Technology Data Exchange (ETDEWEB)

    Zarrabi, K. [New South Wales Univ., Sydney, NSW (Australia). Sch. of Mech. and Mfg. Eng.; Zhang, H. [New South Wales Univ., Sydney, NSW (Australia). Sch. of Mech. and Mfg. Eng.; Nhim, K. [New South Wales Univ., Sydney, NSW (Australia). Sch. of Mech. and Mfg. Eng.

    1997-05-01

    Based on nonlinear finite element analysis, the plastic collapse pressures of cylindrical vessels with longitudinal surface cracks are computed. A general formula of plastic collapse pressure of such structures are given and compared with the literature solutions. The results of the present study could be applied for the integrity assessments, failure analyses, remanent life assessment, and licence extensions of the vessels. (orig.)

  3. 46 CFR 176.812 - Pressure vessels and boilers.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure vessels and boilers. 176.812 Section 176.812... TONS) INSPECTION AND CERTIFICATION Material Inspections § 176.812 Pressure vessels and boilers. (a.... (b) Periodic inspection and testing requirements for boilers are contained in § 61.05 in subchapter F...

  4. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  5. Systems aspects of reactor pressure vessel integrity: Executive report

    International Nuclear Information System (INIS)

    Ray, N.K.; Beck, R.K.; Wallace, I.W.; Grigsby, J.M.; Sloane, B.D.; Bishop, B.A.

    1993-11-01

    Reactor vessel integrity is affected by the following key issues: Operational heatup and cooldown pressure-temperature limit curves and the associated occurrence of low temperature over pressurization (LTOP); pressurized thermal shock (PTS) events that could lead to crack initiation in the vessel, and irradiation effects, including low upper shelf fracture toughness. Electric Power Research Institute's Embrittlement Management Program is underway to develop methods in assessing reactor vessel integrity issues due to radiation embrittlement. Reactor vessel integrity may be impacted by different aspects including some degradation mechanisms. One of those aspects are the plant system changes associated with the reactor vessel. This report uses the results from various probabilistic analyses performed by Westinghouse and other organizations, to identify those plant systems which are most important to the reactor vessel embrittlement issues

  6. Use of L-cysteine for minimization of inorganic Hg loss during thermal neutron irradiation

    International Nuclear Information System (INIS)

    Anderson, D.L.

    2009-01-01

    Thermal neutron irradiation experiments performed with cellulose-based L-cysteine-treated and untreated Hg standards showed Hg losses of 59-81% for untreated standards but only about a 0.2% loss for treated standards. These results and others for multielement standards showed that Hg loss is highly dependent on total mass and placement of materials in the irradiation vessel and that distribution of volatilized Hg was fairly uniform throughout the sample-containing region of the vessel. Polyethylene trapped volatile Hg much more efficiently than cellulose and a multielement standard containing inorganic Se selectively trapped Hg lost from a co-irradiated multielement standard containing Hg. (author)

  7. Neutron irradiation damage in transition metal carbides

    International Nuclear Information System (INIS)

    Matsui, Hisayuki; Nesaki, Kouji; Kiritani, Michio

    1991-01-01

    Effects of neutron irradiation on the physical properties of light transition metal carbides, TiC x , VC x and NbC x , were examined, emphasizing the characterization of irradiation induced defects in the nonstoichiometric composition. TiC x irradiated with 14 MeV (fusion) neutrons showed higher damage rates with increasing C/Ti (x) ratio. A brief discussion is made on 'cascade damage' in TiC x irradiated with fusion neutrons. Two other carbides (VC x and NbC x ) were irradiated with fission reactor neutrons. The irradiation effects on VC x were not so simple, because of the complex irradiation behavior of 'ordered' phases. For instance, complete disordering was revealed in an ordered phase, 'V 8 C 7 ', after an irradiation dose of 10 25 n/m 2 . (orig.)

  8. ATF Neutron Irradiation Program Technical Plan

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post-irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to JAEA in Japan. This report discusses the technical plan of the experimental study.

  9. A neutron irradiator applied to cancer treatment

    International Nuclear Information System (INIS)

    Campos, Tarcisio P.R.; Andrade, Ana P. de

    2000-01-01

    Cancer and the way of treating it with neutron capture therapy are addressed. This paper discusses also the type of neutron facilities used to treat cancer around the world, as follow: discrete neutron sources, accelerators, and nuclear reactors. The major features of an epithermal neutron irradiation facility applied to BNCT treatment are addressed. The main goal is to give another choice of neutron irradiators to be set in a hospital. The irradiation facility embeds a set of 252 Cf neutron source coupled with a homogeneous mixture of uranium-zirconium hydride alloy containing 8.4 wt % uranium enriched to 20% U 235 . The facility delivers an epithermal neutron beam with low background of fast neutron and gamma rays. The N particle transport code (MCNP-4A) has been used during the simulation in order to achieve the desired configurations and to estimate the multiplication factor, k eff . The present facility loaded with 30 mg of 252 Cf neutron source generates an external beam with an intensity of 10 7 n/cm 2 .s on the spectrum of 4 eV to 40 KeV. The 252 Cf - facility coupled with fissile material was able to amplify the epithermal flux to 10 8 n/cm 2 .s, maintaining the figure-of-merits represented by the ratios of the fast dose and gamma dose in air per epithermal neutron flux closed to those values presented by BMRR, MITR-II and Petten Reactor. The medical irradiation facility loaded with 252 Cf- 235 U can be a choice for BNCT. (author)

  10. Atom probe study of the microstructural evolution induced by irradiation in Fe-Cu ferritic alloys and pressure vessel steels; Etude a la sonde atomique de l`evolution microstructurale sous irradiation d`alliages ferritiques Fe-Cu et d`aciers de cuve REP

    Energy Technology Data Exchange (ETDEWEB)

    Pareige, P.

    1996-04-01

    Pressure vessel steels used in pressurized water reactors are low alloyed ferritic steels. They may be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are generally supposed to result from the formation of point defects, dislocation loops, voids and/or copper rich clusters. However, the real nature of the irradiation induced-damage in these steels has not been clearly identified yet. In order to improve our vision of this damage, we have characterized the microstructure of several steels and model alloys irradiated with electrons and neutrons. The study was performed with conventional and tomographic atom probes. The well known importance of the effects of copper upon pressure vessel steel embrittlement has led us to study Fe-Cu binary alloys. We have considered chemical aging as well as aging under electron and neutron irradiations. The resulting effects depend on whether electron or neutron irradiations ar used for thus. We carried out both kinds of irradiation concurrently so as to compare their effects. We have more particularly considered alloys with a low copper supersaturation representative of that met with the French vessel alloys (0.1% Cu). Then, we have examined steels used on French nuclear reactor pressure vessels. To characterize the microstructure of CHOOZ A steel and its evolution when exposed to neutrons, we have studied samples from the reactor surveillance program. The results achieved, especially the characterization of neutron-induced defects have been compared with those for another steel from the surveillance program of Dampierre 2. All the experiment results obtained on model and industrial steels have allowed us to consider an explanation of the way how the defects appear and grow, and to propose reasons for their influence upon steel embrittlement. (author). 3 appends.

  11. Tritium Retention and Permeation in Ion- and Neutron-Irradiated Tungsten under US-Japan PHENIX Collaboration

    Science.gov (United States)

    Shimada, Masashi; Taylor, Chase N.; Kolasinski, Robert D.; Buchenauer, Dean A.; Chikada, Takumi; Oya, Yasuhisa; Hatano, Yuji

    2015-11-01

    A critical challenge for long-term operation of ITER and beyond to a FNSF, a DEMO and future fusion reactor will be the development of plasma-facing components (PFCs) that demonstrate erosion resistance to intense heat and neutral/ion particle fluxes under the extreme fusion nuclear environment, while minimizing in-vessel inventories and ex-vessel permeation of tritium. Recent work at Tritium Plasma Experiment demonstrated that tritium diffuses in bulk tungsten at elevated temperatures, and can be trapped in radiation-induced trap site (up to 1 at. % T/W) in tungsten [M. Shimada, et.al., Nucl. Fusion 55 (2015) 013008]. US-Japan PHENIX collaboration (2013-2019) investigates irradiation response on tritium behavior in tungsten, and performs one-of-a-kind neutron-irradiation with Gd thermal neutron shield at High Flux Isotope Reactor, ORNL. This presentation describes the challenge in elucidating tritium behavior in neutron-irradiated PFCs, the PHENIX plans for neutron-irradiation and post irradiation examination, and the recent findings on tritium retention and permeation in 14MeV neutron-irradiated and Fe ion irradiated tungsten. This work was prepared for the U.S. Department of Energy, Office of Fusion Energy Sciences, under the DOE Idaho Field Office contract number DE-AC07-05ID14517.

  12. Firefighter's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    Beck, E. J.

    1974-01-01

    The research to design, fabricate, test, and deliver a pressure vessel for the main component in an improved high-performance firefighter's breathing system is reported. The principal physical and performance characteristics of the vessel which were required are: (1) maximum weight of 9.0 lb; (2) maximum operating pressure of 4500 psig (charge pressure of 4000 psig); (3) minimum contained volume of 280 in. 3; (4) proof pressure of 6750 psig; (5) minimum burst pressure of 9000 psig following operational and service life; and (6) a minimum service life of 15 years. The vessel developed to fulfill the requirements described was completely sucessful, i.e., every category of performence was satisfied. The average weight of the vessel was found to be about 8.3 lb, well below the 9.0 lb specification requirement.

  13. Seals for sealing a pressure vessel such as a nuclear reactor vessel or the like

    International Nuclear Information System (INIS)

    Bruns, H.J.; Huelsermann, K.H.

    1975-01-01

    A description is given of seals for sealing a pressure vessel such as a nuclear reactor vessel, steam boiler vessel, or any other vessel which is desirably sealed against pressure of the type including a housing and a housing closure that present opposed vertical sealing surfaces which define the sides of a channel. The seals of the present invention comprise at least one sealing member disposed in the channel, having at least one stop face, a base portion and two shank portions extending from the base portion to form a groove-like recess. The shank portions are provided with sealing surfaces arranged to mate with the opposed vertical pressure vessel sealing surfaces. A shank-spreading wedge element also disposed in the channel has at least one stop face and is engaged in the groove-like recess with the sealing member and wedge element stop face adjacent to each other

  14. The need to pressure test prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Forgie, J.H.; Holland, J.A.

    1983-01-01

    In the period when PCRV were relatively unproven, proof pressure testing provided a useful demonstration of vessel integritiy and a confirmation of model testing and of analysis. No failures have occurred during concrete vessel tests in the UK or in the subsequent operational life of the vessels and much has been learned of their behaviour in service. The paper examines the advantages and disadvantages of proof testing PCRV in the light of the above increased knowledge of vessel performance. The paper draws attention to certain hypothetical loading cases that could be more onerous than the proof test and suggests that pressure testing could itself cause unnecessarily high loading to parts of the vessel. Always recognising the safety considerations and demonstrations of such are of prime importance, the authors suggest that a lower pressure level could be adopted without loss of original intent. In addition some ground rules are suggested as to cases where proof testing could be omitted. (orig./HP)

  15. Heating large pressure vessel with a cylindrical cross-section

    International Nuclear Information System (INIS)

    Gabrea, R.; Kaiser, D.; Schwiers, H.G.; Schoening, J.

    1980-01-01

    The proposal concerns the improvement of a construction of the vertical stress system of a prestressed-concrete pressure vessel to take up nuclear power plant components. According to the invention, the vertical bracing, cables are attached outside the pressure-resistant thermal insulation of the pressure container, so that they can be easily checked. They are not subjected to thermal stresses. (UWI) [de

  16. Analysis of properties of WWER-440 reactor pressure vessel semiproducts

    International Nuclear Information System (INIS)

    Horacek, L.; Brumovsky, M.; Brynda, J.

    1988-01-01

    An analysis is made of selected tensile characteristics and of the chemical composition of steels 15Kh2MFA and 18Kh2MFA used for the manufacture of semifinished products of WWER-440 reactor pressure vessels with respect to the first 15 pressure vessels manufactured in Czechoslovakia. The results are analyzed with respect to the developmental trend (comparing the results of analyses of the first 5 and the first 15 pressure vessels manufactured by Skoda) and a different origin (comparing those by Skoda by the Izhorskij Zavod in the USSR). The effects of thickness of the semifinished product on tensile properties and of the chemical composition on radiation resitance of WWER-440 pressure vessels are considered in more detail. (author). 3 tabs., 3 refs

  17. High Toughness Light Weight Pressure Vessel, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Proposed is a pressure vessel with 25% better Fracture Strength over equal strength designed Fiberglass to help reduce 10 to 25% weight for aerospace use. Phase I is...

  18. The intrinsic gettering in neutron irradiation Czochralski-silicon

    CERN Document Server

    Li Yang Xian; Niu Ping Juan; Liu Cai Chi; Xu Yue Sheng; Yang Deren; Que Duan Lin

    2002-01-01

    The intrinsic gettering in neutron irradiated Czochralski-silicon is studied. The result shows that a denuded zone at the surface of the neutron irradiated Czochralski-silicon wafer may be formed through one-step short-time annealing. The width of the denuded zone is dependent on the annealing temperature and the dose of neutron irradiation, while it is irrelated to the annealing time in case the denuded zone is formed. The authors conclude that the interaction between the defects induced by neutron irradiation and the oxygen in the silicon accelerates the oxygen precipitation in the bulk, and becomes the dominating factor of the quick formation of intrinsic gettering. It makes the effect of thermal history as the secondary factor

  19. Variability of mechanical properties of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Petrequin, P.; Soulat, P.

    1980-01-01

    Causes of variability of mechanical properties nuclear pressure vessel steels are reviewed and discussed. The effects of product shape and size, processing history and heat treatment are investigated. Some quantitative informations are given on the scatter of mechanical properties of typical pressure vessel components. The necessity of using recommended or standardized properties for comparing mechanical properties before and after irradiation in pin pointed. (orig.) [de

  20. Adjustable guide for a testing system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Seifert, W.

    1980-01-01

    The device consisting of a guide rail and a manipulator is introduced into the gap between pressure vessel wall and biological shield by means of suspending wire drums and manipulator drums. For adjustment of the device an elbow telescope is used. The guide rail is fixed to the pressure vessel wall by means of electromagnets. The movements of the manipulator with respect to the guide rail are performed with the aid of a motor. (DG) [de

  1. Common pressure vessel development for the nickel hydrogen technology

    Science.gov (United States)

    Holleck, G.

    1981-01-01

    The design of a pressure vessel nickel hydrogen cell is described. The cell has the following key features: it eliminates electrolyte bridging; provides for independent electrolyte management for each unit stack; provides for independent oxygen management for each unit stack; has good heat dissipation; has a mechanically sound and practical interconnection; and has the maximum in common with state of the art individual pressure vessel technology.

  2. Common-Pressure-Vessel Nickel-Hydrogen Battery Development

    OpenAIRE

    Otzinger, Burton; Wheeler, James

    1991-01-01

    The dual-cell, common-pressure vessel, nickel-hydrogen configuration has recently emerged as an option for small satellite nickel-hydrogen battery application. An important incentive is that the dual-cell, CPV configured battery presents a 30 percent reduction in volume and nearly 50 percent reduction in mounting footprint, when compared with an equivalent battery of individual pressure- vessel (IPV) cells. In addition energy density and cost benefits are significant. Eagle-Picher Industries ...

  3. Method of detecting construction faults in concrete pressure vessels

    International Nuclear Information System (INIS)

    Robertson, S.A.; Duhoux, M.; Dawance, G.; Carrie, C.; Morel, D.

    1976-01-01

    A major problem in the design and construction of concrete pressure vessels for nuclear power stations is the risk of excessive air leaks through the concrete itself, due to faulty construction. The 'sonic coring' method of non-destructive concrete testing has been used successfully in pile and diaphragm wall construction control for several years, and the potential use of this method to control the presence of faults in concrete pressure vessels is here described. (author)

  4. Heavy wall pressure vessels for energy systems

    International Nuclear Information System (INIS)

    Canonico, D.A.

    Modifications of steels currently accepted in the Code appear to provide improved mechanical properties. These steels may permit the fabrication of larger diameter vessels with thinner section sizes and improved reliability and integrity. Adapting current specifications should expedite Code approval. Finally the challenge of improving welding procedures and adapting processes for field applications will result in higher quality weldments

  5. Simulation analysis of radiation fields inside phantoms for neutron irradiation

    International Nuclear Information System (INIS)

    Satoh, Daiki; Takahashi, Fumiaki; Endo, Akira; Ohmachi, Y.; Miyahara, N.

    2007-01-01

    Radiation fields inside phantoms have been calculated for neutron irradiation. Particle and heavy-ion transport code system PHITS was employed for the calculation. Energy and size dependences of neutron dose were analyzed using tissue equivalent spheres of different size. A voxel phantom of mouse was developed based on CT images of an 8-week-old male C3H/HeNs mouse. Deposition energy inside the mouse was calculated for 2- and 10-MeV neutron irradiation. (author)

  6. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

  7. In-service supervision of a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Zemann, H.; Mayer, N.; Amberg, C.

    1985-01-01

    On-line measurements of the physical state of a prestressed concrete pressure vessel and a comparison of the distribution of temperature, strain and stress within the concrete member to the optimized statical predictions and the criterions of layout yield to an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed on the prototype vessel at Seibersdorf Research Center during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C/50 bar). (Author)

  8. In-service supervision of a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Zemann, H.; Weissbacher, L.; Mayer, N.; Amberge, C.

    1985-01-01

    On-line measurements of the physical state of a prestressed concrete pressure vessel, and comparison with the design predictions of the distribution of temperature, strain and stress within the concrete member and the criteria of layout, provide an efficient and economical method of operating the vessel with a high potential of safety. The requirements of instrumentation and the comparison with static calculations are discussed with reference to the prototype vessel at Seibersdorf Research Centre during the phase of construction and prestressing, the phase of the first thermal treatment (stabilization), the pressure tests and under the operating conditions of a high temperature reactor (150 0 C, 50 bar). (author)

  9. Phase transformations in neutron-irradiated Zircaloys

    International Nuclear Information System (INIS)

    Chung, H.M.

    1986-04-01

    Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after ∼3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr 3 O and cubic-ZrO 2 particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/,Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/,Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of ∼4 x 10 21 ncm -2 in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs

  10. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  11. Hydriding and neutron irradiation in zircaloy-4

    International Nuclear Information System (INIS)

    Ramos, Ruben Fortunato; Martin, Juan Ezequiel; Orellano, Pablo; Dorao, Carlos; Analia Soldati; Ghilarducci, Ada Albertina; Corso, Hugo Luis; Peretti, Hernan Americo; Bolcich, Juan Carlos

    2003-01-01

    The composition of Zircaloy-4 for nuclear applications is specified by the ASTM B350 Standard, that fixes the amount of alloying elements (Sn, Fe, Cr) and impurities (Ni, Hf, O, N, C, among others) to optimize good corrosion and mechanical behavior.The recycling of zircaloy-4 scrap and chips resulting from cladding tube fabrication is an interesting issue.However, changes in the final composition of the recycled material may occur due to contamination with tool pieces, stainless steel chips, turnings, etc. while scrap is stored and handled. Since the main components of the possible contaminants are Fe, Cr and Ni, it arises the interest in studying up to what limit the Fe, Ni and Cr contents could be exceeded beyond the standard specification without affecting significantly the alloy properties.Zircaloy-4 alloys elaborated with Fe, Cr and Ni additions and others of standard composition in use in nuclear plants are studied by tensile tests, SEM observations and EDS microanalysis.Some samples are tested in the initial condition and others after hydriding treatments and neutron irradiation in the RA6

  12. Effect of neutron irradiation on vanadium alloys

    International Nuclear Information System (INIS)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600 0 C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520 0 C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys

  13. Effect of neutron irradiation on vanadium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Braski, D.N.

    1986-01-01

    Neutron-irradiated vanadium alloys were evaluated for their susceptibility to irradiation hardening, helium embrittlement, swelling, and residual radioactivity, and the results were compared with those for the austenitic and ferritic stainless steels. The VANSTAR-7 and V-15Cr-5Ti alloys showed the greatest hardening between 400 and 600/sup 0/C while V-3Ti-1Si and V-20Ti had lower values that were comparable to those of ferritic steels. The V-15Cr-5Ti and VANSTAR-7 alloys were susceptible to helium embrittlement caused by the combination of weakened grain boundaries and irradiation-hardened grain matrices. Specimen fractures were entirely intergranular in the most severe instances of embrittlement. The V-3Ti-1Si and V-20Ti alloys were more resistant to helium embrittlement. Except for VANSTAR-7 irradiated to 40 dpa at 520/sup 0/C, all of the vanadium alloys exhibited low swelling that was similar to the ferritic steels. Swelling was greater in specimens that were preimplanted with helium using the tritium trick. The vanadium alloys clearly exhibit lower residual radioactivity after irradiation than the ferrous alloys.

  14. Investigation of impulsively loaded pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

    1963-10-15

    Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

  15. Influence of the copper impurity level on the irradiation response of reactor pressure vessel steels investigated by SANS

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, Arne, E-mail: arne.wagner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Safety Research, P.O. Box 510119, 01314 Dresden (Germany); Ulbricht, Andreas; Bergner, Frank; Altstadt, Eberhard [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Safety Research, P.O. Box 510119, 01314 Dresden (Germany)

    2012-06-01

    Reactor pressure vessel (RPV) steel, when exposed to neutron irradiation, induces the formation of nano-sized features. Using small angle neutron scattering (SANS) we have studied the neutron fluence dependence of the precipitate volume fraction for high-Cu and low-Cu materials separately. Cu-rich precipitates have long been recognized to play the dominant role in embrittlement of Cu-bearing RPV steels. In contrast, Mn-Ni-rich precipitates seem to govern embrittlement in the case of low levels of impurity Cu. The objective is to work out the resulting differences from the microstructural point of view. For low-Cu materials, the volume fraction was found to be within the detection limit of SANS at fluences below an apparent threshold fluence, whereas the slope increases considerably beyond. The relationship between irradiation-induced yield stress increase and precipitate volume fraction was also considered. We have derived estimates of the obstacle strength for Cu-rich precipitates and for Mn-Ni-rich precipitates.

  16. Prediction of the shift in the brittle-ductile transition temperature of light-water reactor (LWR) pressure vessel materials

    International Nuclear Information System (INIS)

    Koziol, J.J.

    1983-01-01

    This report presents the results of an investigation undertaken by The Metal Properties Council Subcommittee 6 on Nuclear Materials and ASTM Subcommittee E10.02 on Behavior and Use of Metallic Materials in Nuclear Systems to determine the feasibility of establishing standard design curves for the purpose of predicting changes in the toughness properties of reactor pressure vessel materials as a result of exposure to neutron irradiation. It is based on a statistical treatment of irradiation data available as of November 1977. One of the important products of the MPC-ASTM effort is the computer data bank. It has been carefully scrutinized by members of the MPC and ASTM and its applicability for the present study was unmatched. The periodic updating and expanding the data bank should be considered for future analyses involving revisions to the present study or in conjunction with studies aimed at determining the response of the Charpy upper shelf energy to neutron radiation exposure or both. This report represents the opinion of MPC Subcommittee 6 concerning the predictability of the shift in the brittle-ductile transition temperature based on data available as of November 1977. The data are presented as three sets of curves: total data base, experimental data, and surveillance data

  17. Stress analysis of large pressure vessels

    International Nuclear Information System (INIS)

    Jeanes, P.; King, S.; Wright, M.B.

    1981-01-01

    Five spherical vessels were investigated in this study. Each model had approximately 2,500 20-node solid elements and 12,000 grid points. Three finite element method computer programs were used, MSC/NASTRAN, BERSAFE and FEMALE. The general approach to the problem, the approach to structural modelling, including the shell and nozzles, the stress analysis and the processing of the results are summarized. (U.K.)

  18. Interpretation of Strain Measurements on Nuclear Pressure Vessels

    DEFF Research Database (Denmark)

    Andersen, Svend Ib Smidt; Engbæk, Preben

    1980-01-01

    Selected results from strain measurements on four nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts. The resu......Selected results from strain measurements on four nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzles, internal vessel structure and flange bolts....... The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as a detailed knowledge of the behaviour of the signal...... from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour...

  19. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10 19 n/cm 2 (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10 19 n/cm 2 (>l MeV). In both cases, irradiations were conducted at ∼290 C and annealing treatments were conducted at ∼454 C. The ORNL and RRC

  20. Irradiation, Annealing, and Reirradiation Effects on American and Russian Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Chernobaeva, A.A.; Korolev, Y.N.; Nanstad, R.K.; Nikolaev, Y.A.; Sokolov, M.A.

    1998-06-16

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. Even though a postirradiation anneal may be deemed successful, a critical aspect of continued RPV operation is the rate of embrittlement upon reirradiation. There are insufficient data available to allow for verification of available models of reirradiation embrittlement or for the development of a reliable predictive methodology. This is especially true in the case of fracture toughness data. Under the U.S.-Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS), Working Group 3 on Radiation Embrittlement, Structural Integrity, and Life Extension of Reactor Vessels and Supports agreed to conduct a comparative study of annealing and reirradiation effects on RPV steels. The Working Group agreed that each side would irradiate, anneal, reirradiate (if feasible ), and test two materials of the other. Charpy V-notch (CVN) and tensile specimens were included. Oak Ridge National Laboratory (ORNL) conducted such a program (irradiation and annealing, including static fracture toughness) with two weld metals representative of VVER-440 and VVER-1000 RPVs, while the Russian Research Center-Kurchatov Institute (RRC-KI) conducted a program (irradiation, annealing, reirradiation, and reannealing) with Heavy-Section Steel Technology (HSST) Program Plate 02 and Heavy-Section Steel Irradiation (HSSI) Program Weld 73W. The results for each material from each laboratory are compared with those from the other laboratory. The ORNL experiments with the VVER welds included irradiation to about 1 x 10{sup 19} n/cm{sup 2} (>1 MeV), while the RRC-KI experiments with the U.S. materials included irradiations from about 2 to 18 x 10{sup 19} n/cm{sup 2} (>l MeV). In both cases, irradiations were conducted at {approximately}290 C and annealing treatments were conducted

  1. Transportable, small high-pressure preservation vessel for cells

    International Nuclear Information System (INIS)

    Kamimura, N; Sotome, S; Shimizu, A; Nakajima, K; Yoshimura, Y

    2010-01-01

    We have previously reported that the survival rate of astrocytes increases under high-pressure conditions at 4 0 C. However, pressure vessels generally have numerous problems for use in cell preservation and transportation: (1) they cannot be readily separated from the pressurizing pump in the pressurized state; (2) they are typically heavy and expensive due the use of materials such as stainless steel; and (3) it is difficult to regulate pressurization rate with hand pumps. Therefore, we developed a transportable high-pressure system suitable for cell preservation under high-pressure conditions. This high-pressure vessel has the following characteristics: (1) it can be easily separated from the pressurizing pump due to the use of a cock-type stop valve; (2) it is small and compact, is made of PEEK and weighs less than 200 g; and (3) pressurization rate is regulated by an electric pump instead of a hand pump. Using this transportable high-pressure vessel for cell preservation, we found that astrocytes can survive for 4 days at 1.6 MPa and 4 0 C.

  2. Neutron irradiation of rat embryos in utero

    International Nuclear Information System (INIS)

    Vogel, H.H. Jr.

    1978-01-01

    In the rat radiation is most effective in producing congenital anomalies during the organ-forming period (days 9 to 13), which is approximately equivalent to the 14th to 50th days of human pregnancy. We have exposed female Sprague--Dawley rats on the 18th day of pregnancy to single whole-body doses of fission neutrons (20 to 150 rads). After 20 rads there was a small decrease in body weight which lasted from birth to weaning. During this period 9% of the irradiated rats died compared with 4% of the controls. After 50 rads, 65/275 (23.6%) of the rats died between birth and weaning, and the body-weight loss of the survivors was increased. After 100 rads, 62/133 (47%) died at birth or day 1 and 103/133 (77.4%) died before weaning. A large and significant decrease in body weight persisted in the survivors. After 150 rads of fission neutrons, all 95 rats died within 48 hr of birth. From cross-fostering experiments, we believe this is a direct effect of radiation on the embryos and not an indirect action through the mother or her milk. The LD 50 for the period from birth to weaning is approximately 75 rads of fission neutrons. Studies of organ weight were conducted daily for the first week after birth in an attempt to find the cause of radiation mortality. Body weight of the irradiated animals averaged only about one-half that of the controls. The liver, kidney, brain, and testes of the neutron-irradiated rats weighed significantly less than those of the controls. The weights of the spleen, lungs, duodenum, and stomach were decreased but not significantly. The bone marrow appeared depleted in the irradiated long bones, but the spleen maintained active hematopoiesis 1 to 2 months after neutron exposure

  3. Composite Overwrapped Pressure Vessel (COPV) Stress Rupture Testing

    Science.gov (United States)

    Greene, Nathanael J.; Saulsberry, Regor L.; Leifeste, Mark R.; Yoder, Tommy B.; Keddy, Chris P.; Forth, Scott C.; Russell, Rick W.

    2010-01-01

    This paper reports stress rupture testing of Kevlar(TradeMark) composite overwrapped pressure vessels (COPVs) at NASA White Sands Test Facility. This 6-year test program was part of the larger effort to predict and extend the lifetime of flight vessels. Tests were performed to characterize control parameters for stress rupture testing, and vessel life was predicted by statistical modeling. One highly instrumented 102-cm (40-in.) diameter Kevlar(TradeMark) COPV was tested to failure (burst) as a single-point model verification. Significant data were generated that will enhance development of improved NDE methods and predictive modeling techniques, and thus better address stress rupture and other composite durability concerns that affect pressure vessel safety, reliability and mission assurance.

  4. Dismantling method for reactor pressure vessel and system therefor

    International Nuclear Information System (INIS)

    Hayashi, Makoto; Enomoto, Kunio; Kurosawa, Koichi; Saito, Hideyo.

    1994-01-01

    Upon dismantling of a reactor pressure vessel, a containment building made of concretes is disposed underground and a spent pressure vessel is contained therein, and incore structures are contained in the spent pressure vessel. Further, a plasma-welder and a pressing machine are disposed to a pool for provisionally placing reactor equipments in the reactor building for devoluming the incore structures by welding and compression. An overhead-running crane and rails therefor are disposed on the roof and the outer side of the reactor building for transporting the pressure vessel from the reactor building to the containment building. They may be contained in the containment building after incorporation of the incore structures into the pressure vessel at the outside of the reactor building. For the devoluming treatment, a combination of cutting, welding, pressing and the like are optically conducted. A nuclear power plant can be installed by using a newly manufactured nuclear reactor, with no requirement for a new site and it is unnecessary to provide a new radioactive waste containing facility. (N.H.)

  5. 46 CFR 54.01-17 - Pressure vessel for human occupancy (PVHO).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Pressure vessel for human occupancy (PVHO). 54.01-17... PRESSURE VESSELS General Requirements § 54.01-17 Pressure vessel for human occupancy (PVHO). Pressure vessels for human occupancy (PVHO's) must meet the requirements of subpart B (Commercial Diving Operations...

  6. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54...

  7. Experimental Study of Fast Neutron Irradiation on Si Transistor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sung Ho; Sun, Gwang Min; Baek, Ha ni; Jin, Seong Bok; Hoang, Sy Minh Tuan [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Bipolar junction transistors (BJTs) are applied in many industrial fields. BJT is a three-terminal device with an important feature in that the current through two terminals can be controlled by small changes we make in the current or voltage at the third terminal. This control feature allows us to amplify small AC signals or to switch the device from an on state and off state and back. These two operations, amplification and switching, are the basis of a host of electronic functions. This study will investigate the electrical characteristics of a p-n-p BJT, such as the base current and collector current for fast neutron irradiation. Fast neutron irradiation can cause displacement damage in the Si bulk. In this paper, the electrical characteristics of a p-n-p BJT such as a base current and collector current are investigated for fast neutron irradiation. The experimental results show that the base current is increased and the collector current is decreased after fast neutron irradiation. These results indicate that the displacement damage caused by fast neutron irradiation increases the recombination rate of minority carriers and resistors.

  8. Pressure vessel burst test program - Progress paper No. 3

    Science.gov (United States)

    Cain, Maurice R.; Sharp, Douglas E.; Coleman, Michael D.

    1992-01-01

    An updated progress report is provided on a program developed to study through test and analysis, the characteristics of blast waves and fragmentation generated by ruptured gas filled pressure vessels. Prior papers on this USAF/NASA/General Physics program were presented to the AIAA in July 1990 and June 1991. Ten pressure vessels have been burst using pneumatic pressure. Tests were designed to explore burst characteristics and used an instrumented arena. Data trends for current experiments are presented. This paper is the third progress report on the program and addresses: (1) a brief review of current methods for assessing vessel safety and burst parameters, (2) a review of pneumatic burst testing operations and testing results, including a comparison to current methods for burst assessment, and (3) a review of the basis for the current test program including planned testing.

  9. Cooling device for upper lid of reactor pressure vessel

    International Nuclear Information System (INIS)

    Takayama, Kazuhiko.

    1997-01-01

    An upper lid of a reactor pressure vessel in a BWR type reactor has one or more annular cooling elements on the surface. The outer side thereof is covered by a temperature keeping frame. The cooling elements have an annular hollow shape, and cooling water is supplied to the hollow portion. As the cooling water supplied to the cooling elements, cooling water of a reactor auxiliary cooling system as a cooling water system incorporated in the reactor container or cooling water of a dehumidification system of the reactor container is used. A plurality of temperature sensors are disposed to various portions of the upper lid of the pressure vessel. A control device determines scattering of the temperature for the entire upper lid of the pressure vessel by temperature signals sent from the temperature sensors. Then, the amount of cooling water to be flown to each of the cooling elements is controlled so as to eliminate the scattering. (I.N.)

  10. The Assembly and Test of Pressure Vessel for Irradiation

    International Nuclear Information System (INIS)

    Park, Kook Nam; Lee, Jong Min; Youn, Young Jung; June, Hyung Kil; Ahn, Sung Ho; Lee, Kee Hong; Kim, Young Ki; Kennedy, Timothy C.

    2009-01-01

    The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts: the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature

  11. Industrial safety of pressure vessels - structural integrity point of view

    Directory of Open Access Journals (Sweden)

    Sedmak Aleksandar

    2016-01-01

    Full Text Available This paper presents different aspects of pressure vessel safety in the scope of industrial safety, focused to the chemical industry. Quality assurance, including application of PED97/23 has been analysed first, followed shortly by the risk assessment and in details by the structural integrity approach, which has been illustrated with three case studies. One important conclusion, following such an approach, is that so-called water proof testing can actually jeopardize integrity of a pressure vessel instead of proving it. [Projekat Ministarstva nauke Republike Srbije, br. TR 174004 i br. TR 33044

  12. A prototype knowledge based system for pressure vessel design

    Energy Technology Data Exchange (ETDEWEB)

    Gunnarsson, L.

    1991-11-22

    The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD`s language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

  13. A prototype knowledge based system for pressure vessel design

    Energy Technology Data Exchange (ETDEWEB)

    Gunnarsson, L.

    1991-11-22

    The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au).

  14. Application of fracture mechanics to fatigue in pressure vessels

    International Nuclear Information System (INIS)

    Ghavami, K.

    1982-01-01

    The methods of application of fracture mechanics to predict fatigue crack propagation in welded structures and pressure vessels are described with the following objectives: i) To identify the effect of different variables such as crack tip plasticity, free surface, finite plate thickness, stress concentration and type of the structure, on the magnitude of stress intensity factor K in Welded joint. ii) To demonstrate the use of fracture mechanics for analysing fatigue crack propagation data. iii) To show how a law of fatigue crack propagation based on fracure mechanics, may be used to predict fatigue behavior of welded structures such as pressure vessel. (Author) [pt

  15. Cylindrical prestressed concrete pressure vessel for a nuclear power plant

    International Nuclear Information System (INIS)

    Horner, M.; Hodzic, A.; Haferkamp, D.

    1976-01-01

    A prestressed concrete pressure vessel for a HTGR is proposed which encloses, in addition to the reactor core, not only the heat-exchanging facilities but also the turbine unit. The reinforcement of the cylindrical concrete body is to be carried out with special care, it is provided for horizontal tendons, the prestressed concrete pressure vessel has a wire-winding device, while the longitudinal reinforcement is achieved by tendous guided in parallel to the vesses axes through the interspaces between the pods. (UWI) [de

  16. A prototype knowledge based system for pressure vessel design

    International Nuclear Information System (INIS)

    Gunnarsson, L.

    1991-01-01

    The usage of expert system techniques in the area of mechanical engineering design has been studied. A prototype expert system for pressure vessel design has been developed. The work has been carried out in two steps. Firstly, a pre-processor for the finite element system PCFEMP, named INFEMP, was developed. Secondly, an expert supported system for pressure vessel design, named PVES, was developed. Both INFEMP and PVES are integrated to the AutoCAD system, and AutoCAD's language AutoLISP has been used. A practical example has been investigated to demonstrate the principal ideas of the prototype. (au)

  17. Reflection and photoemission studies of neutron-irradiated graphite

    International Nuclear Information System (INIS)

    Fukutani, Hirohito; Yamada, Akio; Yagi, Kazutoshi; Ooe, Satoshi; Higashiyama, Kazuyuki; Kato, Hiroo; Iwata, Tadao.

    1990-01-01

    Neutron-irradiated graphites were studied by reflectivity and photoemission (UPS, ARUPS, XPS) measurements. The π-band reflectivity peak of graphite, located at 5 eV, changed significantly and a small absorption band ascribed to vacancies produced by neutron bombardment was found to grow around 3 eV. Modification of the valence band by neutron irradiation was studied by ARUPS. The π-valence band shifts to lower binding energy towards the Fermi level and its band width becomes smaller. These results were also confirmed by the optical joint density of states obtained from K-K analysis of the reflectivity. (author)

  18. Behavior of fluorine 18 in neutron irradiated zeolites

    International Nuclear Information System (INIS)

    Estevez Lopez, D.R.

    1992-01-01

    The transformation of Li-exchanged H-Y zeolite has been investigated at 300, 550, 850 and 1050 Centigrade degree, formation of quartz structure in addition to an amorphous phase, was nited. The Li-aluminosilicate obtained was neutron irradiated and the chemical behavior of 18 F produced by the reaction sequence 6 Li (n, α) 3 H, 16 O ( 3 H, n) 18 F, was studied. The neutron irradiated material was purged with argon-hydron gas streams. It was found that the amount of released 18 F depends on the temperature used (Author)

  19. Heritability of retinal vessel diameters and blood pressure

    DEFF Research Database (Denmark)

    Taarnhøj, Nina C B B; Larsen, Michael; Sander, Birgit

    2006-01-01

    for CRVE, and 0.67 +/- 0.05 microm for AVR. No significant influence on artery or vein diameters was found for gender, smoking, body mass index (BMI), total cholesterol, fasting blood glucose, or 2-hour oral glucose tolerance test values. CONCLUSIONS: In healthy young adults with normal blood pressure......PURPOSE: To assess the relative influence of genetic and environmental effects on retinal vessel diameters and blood pressure in healthy adults, as well as the possible genetic connection between these two characteristics. METHODS: In 55 monozygotic and 50 dizygotic same-sex healthy twin pairs......%-80%) for CRAE, 83% (95% CI: 73%-89%) for CRVE, and 61% (95% CI: 44%-73%) for mean arterial blood pressure (MABP). Retinal artery diameter decreased with increasing age and increasing arterial blood pressure. Mean vessel diameters in the population were 165.8 +/- 14.9 microm for CRAE, 246.2 +/- 17.7 microm...

  20. Reliability aspects of radiation damage in reactor pressure vessel mterials

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1985-01-01

    The service life estimate is a major factor in the evaluation of the operating reliability and safety of a nuclear reactor pressure vessel. The evaluation of the service life of the pressure vessel is based on a comparison of fracture toughness values with stress intensity factors. Notch toughness curves are used for the indirect determination of fracture toughness. The dominant degradation effect is radiation embrittlement. Factors having the greatest effect on the result are the properties of the starting material of the vessel and the impurity content, mainly the Cu and P content. The design life is affected by the evaluation of residual lifetime which is made by periodical nondestructive inspections and using surveillance samples. (M.D.)

  1. Possible research program on a large scale nuclear pressure vessel

    International Nuclear Information System (INIS)

    1983-01-01

    The nuclear pressure vessel structural integrity is actually one of the main items in the nuclear plants safety field. An international study group aimed at investigating the feasibility of a ''possible research program'' on a scale 1:1 LWR pressure vessel. This report presents the study group's work. The different research programs carried out or being carried out in various countries of the European Community are presented (phase I of the study). The main characteristics of the vessel considered for the program and an evaluation of activities required for making them available are listed. Research topic priorities from the different interested countries are summarized in tables (phase 2); a critical review by the study group of the topic is presented. Then, proposals for possible experimental programs and combination of these programs are presented, only as examples of possible useful research activities. The documents pertaining to the results of phase I inquiry performed by the study group are reported in the appendix

  2. Multilayer Pressure Vessel Materials Testing and Analysis. Phase 1

    Science.gov (United States)

    Cardinal, Joseph W.; Popelar, Carl F.; Page, Richard A.

    2014-01-01

    To provide NASA a comprehensive suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for aging multilayer pressure vessels, Southwest Research Institute (R) (SwRI) was contracted in two phases to obtain relevant material property data from a representative vessel. This report describes Phase 1 of this effort which includes a preliminary material property assessment as well as a fractographic, fracture mechanics and fatigue crack growth analyses of an induced flaw in the outer shell of a representative multilayer vessel that was subjected to cyclic pressure test. SwRI performed this Phase 1 effort under contract to the Digital Wave Corporation in support of their contract to Jacobs ATOM for the NASA Ames Research Center.

  3. Long-term irradiation effects on reactor-pressure vessel steels. Investigations on the nanometer scale

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, Arne

    2017-06-01

    The exposure of reactor pressure vessel (RPV) steels to neutron irradiation gives rise to irradiation-enhanced diffusion, a rearrangement of solute atoms and, consequently, a degradation of the mechanical properties. The increasing age of existing nuclear power plants raises new questions specific to long-term operation. Two of them are addressed in this thesis: flux effects and the late-blooming effect. Can low-flux irradiations up to a given fluence be reproduced by more rapid high-flux irradiations up to the same fluence? Can the irradiation response of RPV steels be extrapolated to higher fluences or are there unexpected ''late-blooming'' effects. Small-angle neutron scattering (SANS), atom-probe tomography (APT) and Vickers-hardness testing were applied. A novel Monte-Carlo based fitting algorithm for SANS data was implemented in order to derive statistically reliable characteristics of irradiation-induced solute-atom clusters. APT was applied in selected cases to gain additional information on the composition and the shape of clusters. Vickers hardness testing was performed on the SANS samples to link the nanometer-scale changes to irradiation hardening. The investigations on flux effects show that clusters forming upon high-flux irradiation are smaller and tend to have a higher number density compared to low-flux irradiations at a given neutron fluence. The measured flux dependence of the cluster-size distribution is consistent with the framework of deterministic growth (but not with coarsening) in combination with radiation-enhanced diffusion. Since the two effects on cluster-size and volume fraction partly cancel each other out, no significant effect on the hardening is observed. The investigations of a possible late-blooming effect indicate that the very existence (yes or no) of such an effect depends on the irradiation conditions. Irradiations at lower fluxes and a lower temperature (255 C) give rise to a significant increase of the

  4. Device for the burst protection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Daublebsky, P.

    1976-01-01

    The burst protection device has a hood over top and bottom of the pressure vessel with superimposed hinged supports lying in their turn against supporting rings which are connected with each other by vertical bracing. It is proposed to place an intermediate layer between hoods and vertical bracing absorbing thermal stresses, i.e. deforming plastically with gradually increasing pressure, but behaving like a rigid body in the case of shock loads. As a material lead e.g. is proposed. (UWI) [de

  5. Why and how acoustic emission in pressure vessel first hydrotest

    International Nuclear Information System (INIS)

    Panzani, C.; Tonolini, F.; Villa, G.; Regis, V.

    1985-01-01

    The main advantages obtained performing the Acoustic Emission (AE) examination during pressure vessel first hydrotest are presented. The characteristics and performance of the AE instrumentation to be used for a correct test are illustrated. The main criteria for AE source characterization (location, typical AE parameters and their correlation with pressure value), the calibration and test procedures are discussed. The ndt post-test examinations and laboratory specimen experiments are also outlined. Personnel qualification requirements are finally indicated. (Author) [pt

  6. Recent experiences and problems in conducting pressure vessel surveillance examinations

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1979-01-01

    Each of the commercial power reactors in the U.S.A. has a pressure vessel surveillance program. The purpose of the programs is to monitor the effects of radiation on the mechanical properties on the steel pressure vessels. A program for a given reactor includes a series of irradiation capsules containing neutron dosimeters and mechanical property specimens. The capsules are periodically removed during the life of the reactor and evaluated. The surveillance capsule examinations conducted to date have been valuable in assessing the effects of radiation on pressure vessels. However, a number of problems have been observed in the course of capsule examinations which potentially could reduce the maximum value of the data obtained. These problems are related to specimen design and preparation, capsule design and preparation, capsule installation and removal, capsule disassembly, specimen testing and evaluation, program documentation, and quality assurance. Examples of problems encountered in the preceding areas are presented in the present paper, and recommendations are made for minimization or prevention of these problems in future programs. Included in the recommendations is that appropriate ASTM standards, ASME Boiler and Pressure Vessel Code sections, and NRC regulations provide the appropriate framework for prevention of problems

  7. Requirements on crack detection in pressurized vessels for ASME authorization

    International Nuclear Information System (INIS)

    Spremo, N.; Begovich, B.; Doko, A.

    1986-01-01

    The method is presented of training qualified personnel for nondestructive testing of pressure vessels. Personnel is divided according to experience and previous training into three groups of which each has its own educational programme. Written examinations in general knowledge and in specialized subjects and a practical examination in crack detection terminate the training. (E.S.). 1 fig., 4 tabs., 3 refs

  8. Design and Optimization of Filament Wound Composite Pressure Vessels

    NARCIS (Netherlands)

    Zu, L.

    2012-01-01

    One of the most important issues for the design of filament-wound pressure vessels reflects on the determination of the most efficient meridian profiles and related fiber architectures, leading to optimal structural performance. To better understand the design and optimization of filament-wound

  9. Initiation and arrest - two approaches to pressure vessel safety

    International Nuclear Information System (INIS)

    Brumovsky, M.; Filip, R.; Stepanek, S.

    1976-01-01

    The safety analysis is described of the reactor pressure vessel related to brittle fracture based on the fracture mechanics theory using two different approximations, i.e., the Crack Arrest Temperature (CAT) or Nil Ductility Temperature (NDT), and fracture toughness. The variation of CAT with stress was determined for different steel specimens of 120 to 200 mm in thickness. A diagram is shown of CAT variation with stress allowing the determination of crack arrest temperature for all types of commonly used steels independently of the NDT initial value. The diagram also shows that the difference between fracture transition elastic (FTE) and NDT depends on the type of material and determines the value of the ΔTsub(sigma) factor typical of the safety coefficient. The so-called fracture toughness reference value Ksub(IR) is recommended for the computation of pressure vessel criticality. Also shown is a defect analysis diagram which may be used for the calculation of pressure vessel safety prior to and during operation and which may also be used in making the decision on what crack sizes are critical, what cracks may be arrested and what cracks are likely to expand. The diagram is also important for the fact that it is material-independent and may be employed for the estimates of pre-operational and operational inspections and for pressure vessel life prediction. It is generally applicable to materials of greater thickness in the region where the validity of linear elastic fracture mechanics is guaranteed. (J.P.)

  10. USER SPECIFICATIONS FOR PRESSURE VESSELS AND TECHNICAL INTEGRITY

    Directory of Open Access Journals (Sweden)

    K.S. Johnston

    2012-01-01

    Full Text Available

    ENGLISH ABSTRACT: Specifications translated from user requirements are prescribed in an attempt to capture and incorporate best practices with regards to the design, fabrication, testing, and operation of pressure vessels. The question as to whether these requirements affect the technical integrity of pressure vessels is often a subjective matter. This paper examines typical user requirement specifications against technical integrity of pressure vessels.
    The paper draws on a survey of a convenience sample of practising engineers in a diversified petrochemical company. When compared with failures on selected pressure vessels recorded by Phillips and Warwick, the respondent feedback confirms the user specifications that have the highest impact on technical integrity.

    AFRIKAANSE OPSOMMING: Gebruikersbehoeftes word saamgevat in spesifikasies wat lei tot goeie praktyk vir ontwerp, vervaarding, toetsing en bedryf van drukvate. Subjektiwiteit van die gebruikersbehoeftes mag soms die tegniese integriteit van ‘n drukvat beinvloed.
    Die navorsing maak by wyse van monsterneming gebruik van die kennis van ingenieurs wat werk in ‘n gediversifiseerde petrochemiese bedryf. Die terugvoering bevestig dat bogenoemde spesifikasies inderdaad die grootste invloed het op tegniese integriteit.

  11. Fiber Optic Acoustic Emission system for structural health monitoring of Composite Pressure Vessels, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Pressurized systems and pressure vessels used in NASA ground-based and flight-based applications including fuel tanks, composite overwrapped pressure vessels...

  12. 46 CFR 97.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 97.30-1 Section... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 97.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer...

  13. 46 CFR 196.30-1 - Repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Repairs to boilers and pressure vessels. 196.30-1... VESSELS OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 196.30-1 Repairs to boilers and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the Chief Engineer...

  14. Emulation of neutron irradiation effects with protons: validation of principle

    International Nuclear Information System (INIS)

    Was, G.S.; Busby, J.T.; Allen, T.; Kenik, E.A.; Jensson, A.; Bruemmer, S.M.; Gan, J.; Edwards, A.D.; Scott, P.M.; Andreson, P.L.

    2002-01-01

    This paper presents the results of the irradiation, characterization and irradiation assisted stress corrosion cracking (IASCC) behavior of proton- and neutron-irradiated samples of 304SS and 316SS from the same heats. The objective of the study was to determine whether proton irradiation does indeed emulate the full range of effects of in-reactor neutron irradiation: radiation-induced segregation (RIS), irradiated microstructure, radiation hardening and IASCC susceptibility. The work focused on commercial heats of 304 stainless steel (heat B) and 316 stainless steel (heat P). Irradiation with protons was conducted at 360 deg. C to doses between 0.3 and 5.0 dpa to approximate those by neutron irradiation at 275 deg. C over the same dose range. Characterization consisted of grain boundary microchemistry, dislocation loop microstructure, hardness as well as stress corrosion cracking (SCC) susceptibility of both un-irradiated and irradiated samples in oxygenated and de-oxygenated water environments at 288 deg. C. Overall, microchemistry, microstructure, hardening and SCC behavior of proton- and neutron-irradiated samples were in excellent agreement. RIS analysis showed that in both heats and for both irradiating particles, the pre-existing grain boundary Cr enrichment transformed into a 'W' shaped profile at 1.0 dpa and then into a 'V' shaped profile between 3.0 and 5.0 dpa. Grain boundary segregation of Cr, Ni, Si, and Mo all followed the same trends and agreed well in magnitude. The microstructure of both proton- and neutron-irradiated samples was dominated by small, faulted dislocation loops. Loop size distributions were nearly identical in both heats over a range of doses. Saturated loop size following neutron irradiation was about 30% larger than that following proton irradiation. Loop density increased with dose through 5.0 dpa for both particle irradiations and was a factor of 3 greater in neutron-irradiated samples vs. proton-irradiated samples. Grain boundary

  15. Residual Stress Estimation and Fatigue Life Prediction of an Autofrettaged Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyung Jin; Kim, Eun Kyum; Koh, Seung Kee [Kunsan Nat’l Univ., Kunsan (Korea, Republic of)

    2017-09-15

    Fatigue failure of an autofrettaged pressure vessel with a groove at the outside surface occurs owing to the fatigue crack initiation and propagation at the groove root. In order to predict the fatigue life of the autofrettaged pressure vessel, residual stresses in the autofrettaged pressure vessel were evaluated using the finite element method, and the fatigue properties of the pressure vessel steel were obtained from the fatigue tests. Fatigue life of a pressure vessel obtained through summation of the crack initiation and propagation lives was calculated to be 2,598 cycles for an 80% autofrettaged pressure vessel subjected to a pulsating internal pressure of 424 MPa.

  16. Advanced Approach of Reactor Pressure Vessel In-service Inspection

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Pajnic, M.

    2006-01-01

    The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

  17. Stress analysis in a non axisymmetric loaded reactor pressure vessel

    International Nuclear Information System (INIS)

    Albuquerque, Levi Barcelos; Assis, Gracia Menezes V. de; Miranda, Carlos Alexandre J.; Cruz, Julio Ricardo B.; Mattar Neto, Miguel

    1995-01-01

    In this work we intend to present the stress analysis of a PWR vessel under postulated concentrated loads. The vessel was modeled with Axisymmetric solid 4 nodes harmonic finite elements with the use of the ANSYS program, version 5.0. The bolts connecting the vessel flanges were modeled with beam elements. Some considerations were made to model the contact between the flanges. The perforated part of the vessel tori spherical head was modeled (with reduced properties due to its holes) to introduce its stiffness and loads but was not within the scope of this work. The loading consists of some usual ones, as pressure, dead weight, bolts preload, seismic load and some postulated ones as concentrated loads, over the vessel, modeled by Fourier Series. The results in the axisymmetric model are taken in terms of linearized stresses, obtained in some circumferential positions and for each position, in some sections along the vessel. Using the ASME Code (Section III, Division 1, Sub-section NB) the stresses are within the allowable limits. In order to draw some conclusions about stress linearization, the membrane plus bending stresses (Pl + Pb) are obtained and compared in some sections, using three different methods. (author)

  18. Circular cylinders and pressure vessels stress analysis and design

    CERN Document Server

    Vullo, Vincenzo

    2014-01-01

    This book provides comprehensive coverage of stress and strain analysis of circular cylinders and pressure vessels, one of the classic topics of machine design theory and methodology. Whereas other books offer only a partial treatment of the subject and frequently consider stress analysis solely in the elastic field, Circular Cylinders and Pressure Vessels broadens the design horizons, analyzing theoretically what happens at pressures that stress the material beyond its yield point and at thermal loads that give rise to creep. The consideration of both traditional and advanced topics ensures that the book will be of value for a broad spectrum of readers, including students in postgraduate, and doctoral programs and established researchers and design engineers. The relations provided will serve as a sound basis for the design of products that are safe, technologically sophisticated, and compliant with standards and codes and for the development of innovative applications.

  19. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  20. Improved Attachment in a Hybrid Inflatable Pressure Vessel

    Science.gov (United States)

    Johnson, Christopher J.; Patterson, Ross; Spexarth, Gary R.

    2010-01-01

    The vessel is a hybrid that comprises an inflatable shell attached to a rigid structure. The inflatable shell is, itself, a hybrid that comprises (1) a pressure bladder restrained against expansion by (2) a restraint layer that comprises a web of straps made from high-strength polymeric fabrics. The present improvements are intended to overcome deficiencies in those aspects of the original design that pertain to attachment of the inflatable shell to the rigid structure. In a typical intended application, such attachment(s) would be made at one or more window or hatch frames to incorporate the windows or hatches as integral parts of the overall vessel.

  1. Evaluation of gamma and neutron irradiation effects on the ...

    Indian Academy of Sciences (India)

    Unknown

    Abstract. We present an investigation of gamma and neutron radiation effects on mica film capacitors from an electrical point of view. We have studied quantitatively the effects of gamma and neutron irradiation on mica film capacitors of thickness, 20 and 40 µm (0⋅7874 and 1⋅5748 mil) with two different areas, 01 and.

  2. Evaluation of gamma and neutron irradiation effects on the ...

    Indian Academy of Sciences (India)

    We present an investigation of gamma and neutron radiation effects on mica film capacitors from an electrical point of view. We have studied quantitatively the effects of gamma and neutron irradiation on mica film capacitors of thickness, 20 and 40 m (0.7874 and 1.5748 mil) with two different areas, 01 and 04 cm2.

  3. Stable and unstable crack growth in pressure vessel models

    International Nuclear Information System (INIS)

    Smith, G.C.; Canonico, D.A.; Stelzman, W.J.

    1978-01-01

    Three identical steel pressure vessels with 254-mm (10-in.) dia, 38-mm (1.5-in.) wall thicknesses and long, deep machined and sharpened axially oriented flaws were tested at three different temperatures. The vessels were assembled by electron-beam welding cylindrical sections with substantially different toughnesses due to different heat treatments. Crack extension initiated in relatively brittle sections, and the cracks extended both stably and unstably, depending on test temperature, toward the tougher sections where crack arrest did and did not occur. Charpy impact specimens and both slow-bend and dynamic precracked Charpy specimens were used for material characterization. The behavior of the vessels is described and related to the Charpy data

  4. Mechanical probabilistic study of pressurized water reactor (PWR) vessels

    International Nuclear Information System (INIS)

    Venturini, V.; Persoz, M.; Debost-Eymard; Donore, A.M.; Meister, E.

    1999-01-01

    This paper describes the probabilistic study carried out on the vessel case under assumptions of occurrence of an underclad flaw and of an accidental transient of the pressurized thermal shock type (PTS). The estimation model of the failure criterion requires finite element computations. Therefore the probabilistic analysis is based on a coupling between Code Aster, the EDF FE software for thermomechanical analysis and PROBAN, a probabilistic analysis software (Det Norske Veritas). The vessel model is first described by focusing on the thermomechanical model (pseudo-1D and 2D axisymmetrical computations are considered and compared) then on the model used in the probabilistic analysis. The second part deals with the probabilistic analyses (computations of failure probabilities during the transient) carried out on the model for different cases and the comparison of the results. Eventually, the perspectives of this work, from both the computational point of view and the application to the vessel study, are outlined. (orig./DGE)

  5. 46 CFR 109.421 - Report of repairs to boilers and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 4 2010-10-01 2010-10-01 false Report of repairs to boilers and pressure vessels. 109... Report of repairs to boilers and pressure vessels. Before making repairs, except normal repairs and maintenance such as replacement of valves or pressure seals, to boilers or unfired pressure vessels in...

  6. 29 CFR 1915.172 - Portable air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 7 2010-07-01 2010-07-01 false Portable air receivers and other unfired pressure vessels... SHIPYARD EMPLOYMENT Portable, Unfired Pressure Vessels, Drums and Containers, Other Than Ship's Equipment § 1915.172 Portable air receivers and other unfired pressure vessels. (a) Portable, unfired pressure...

  7. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  8. Rupture tests with reactor pressure vessel head models

    International Nuclear Information System (INIS)

    Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

    2003-01-01

    In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

  9. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifting Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, S. L.

    2009-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle s Kevlar-49 (DuPont) fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed nonconservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23 percent lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  10. Composite Overwrap Pressure Vessels: Mechanics and Stress Rupture Lifing Philosophy

    Science.gov (United States)

    Thesken, John C.; Murthy, Pappu L. N.; Phoenix, Leigh

    2007-01-01

    The NASA Engineering and Safety Center (NESC) has been conducting an independent technical assessment to address safety concerns related to the known stress rupture failure mode of filament wound pressure vessels in use on Shuttle and the International Space Station. The Shuttle's Kevlar-49 fiber overwrapped tanks are of particular concern due to their long usage and the poorly understood stress rupture process in Kevlar-49 filaments. Existing long term data show that the rupture process is a function of stress, temperature and time. However due to the presence of load sharing liners and the complex manufacturing procedures, the state of actual fiber stress in flight hardware and test articles is not clearly known. Indeed non-conservative life predictions have been made where stress rupture data and lifing procedures have ignored the contribution of the liner in favor of applied pressure as the controlling load parameter. With the aid of analytical and finite element results, this paper examines the fundamental mechanical response of composite overwrapped pressure vessels including the influence of elastic-plastic liners and degraded/creeping overwrap properties. Graphical methods are presented describing the non-linear relationship of applied pressure to Kevlar-49 fiber stress/strain during manufacturing, operations and burst loadings. These are applied to experimental measurements made on a variety of vessel systems to demonstrate the correct calibration of fiber stress as a function of pressure. Applying this analysis to the actual qualification burst data for Shuttle flight hardware revealed that the nominal fiber stress at burst was in some cases 23% lower than what had previously been used to predict stress rupture life. These results motivate a detailed discussion of the appropriate stress rupture lifing philosophy for COPVs including the correct transference of stress rupture life data between dissimilar vessels and test articles.

  11. Quality Testing of Gaseous Helium Pressure Vessels by Acoustic Emission

    CERN Document Server

    Barranco-Luque, M; Hervé, C; Margaroli, C; Sergo, V

    1998-01-01

    The resistance of pressure equipment is currently tested, before commissioning or at periodic maintenance, by means of normal pressure tests. Defects occurring inside materials during the execution of these tests or not seen by usual non-destructive techniques can remain as undetected potential sources of failure . The acoustic emission (AE) technique can detect and monitor the evolution of such failures. Industrial-size helium cryogenic systems employ cryogens often stored in gaseous form under pressure at ambient temperature. Standard initial and periodic pressure testing imposes operational constraints which other complementary testing methods, such as AE, could significantly alleviate. Recent reception testing of 250 m3 GHe storage vessels with a design pressure of 2.2 MPa for the LEP and LHC cryogenic systems has implemented AE with the above-mentioned aims.

  12. Stress Rupture Life Reliability Measures for Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Murthy, Pappu L. N.; Thesken, John C.; Phoenix, S. Leigh; Grimes-Ledesma, Lorie

    2007-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are often used for storing pressurant gases onboard spacecraft. Kevlar (DuPont), glass, carbon and other more recent fibers have all been used as overwraps. Due to the fact that overwraps are subjected to sustained loads for an extended period during a mission, stress rupture failure is a major concern. It is therefore important to ascertain the reliability of these vessels by analysis, since the testing of each flight design cannot be completed on a practical time scale. The present paper examines specifically a Weibull statistics based stress rupture model and considers the various uncertainties associated with the model parameters. The paper also examines several reliability estimate measures that would be of use for the purpose of recertification and for qualifying flight worthiness of these vessels. Specifically, deterministic values for a point estimate, mean estimate and 90/95 percent confidence estimates of the reliability are all examined for a typical flight quality vessel under constant stress. The mean and the 90/95 percent confidence estimates are computed using Monte-Carlo simulation techniques by assuming distribution statistics of model parameters based also on simulation and on the available data, especially the sample sizes represented in the data. The data for the stress rupture model are obtained from the Lawrence Livermore National Laboratories (LLNL) stress rupture testing program, carried out for the past 35 years. Deterministic as well as probabilistic sensitivities are examined.

  13. Niobium Application, Metallurgy and Global Trends in Pressure Vessel Steels

    Science.gov (United States)

    Jansto, Steven G.

    Niobium-containing high strength steel materials have been developed for a variety of pressure vessel applications. Through the application of these Nb-bearing steels in demanding applications, the designer and end user experience improved toughness at low temperature, excellent fatigue resistance and fracture toughness and excellent weldability. These enhancements provide structural engineers the opportunity to further improve the pressure vessel design and performance. The Nb-microalloy alloy designs also result in reduced operational production cost at the steel operation, thereby embracing the value-added attribute Nb provides to both the producer and the end user throughout the supply chain. For example, through the adoption of these Nb-containing structural materials, several design-manufacturing companies are considering improved designs which offer improved manufacturability, lower overall cost and better life cycle performance.

  14. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  15. Backstreaming of Impurity Gas Through a Leak in Pressurized Vessel

    CERN Document Server

    Dauvergne, J P

    1998-01-01

    The presence of a leak in a vessel containing pure gas can induce the contamination by atmospheric gas diffusing into the vessel. In order to avoid this, a gas which has to be kept pure also in presen ce of a leak is usually pressurized, to reduce the flow of contaminating gas through the leak owing to the molecular drag by the outstreaming pure gas. In this paper, a simple model calculation of ba ckstreaming based on the solution of the diffusion + drag equation in cylindrical coordinates is presented. It is shown that both the pressure difference and the dimension of the leak are critical in determining the contaminating flow, a maximum in the backstreaming flow appearing when the drag velocity of the outstreaming gas equals the diffusion velocity.

  16. Evaluation of Agency Non-Code Layered Pressure Vessels (LPVs)

    Science.gov (United States)

    Prosser, William H.

    2014-01-01

    In coordination with the Office of Safety and Mission Assurance and the respective Center Pressure System Managers (PSMs), the NASA Engineering and Safety Center (NESC) was requested to formulate a consensus draft proposal for the development of additional testing and analysis methods to establish the technical validity, and any limitation thereof, for the continued safe operation of facility non-code layered pressure vessels. The PSMs from each NASA Center were asked to participate as part of the assessment team by providing, collecting, and reviewing data regarding current operations of these vessels. This report contains the outcome of the assessment and the findings, observations, and NESC recommendations to the Agency and individual NASA Centers.

  17. A classification system for pressure vessel shell failures

    International Nuclear Information System (INIS)

    Harrop, L.P.

    1989-01-01

    A system for classifying failures of the shells of pressure vessels is presented. The classification system is based on the way a failure physically manifests itself and not on imputed economic or safety significance. It is believed the described way of classifying the failures is useful for transferring information from one situation to another. In assigning names to types of failure, the intention has been to adopt explicit definitions rather than supposed colloquial usage. (author)

  18. H.B. Robinson-2 pressure vessel benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Remec, I.; Kam, F.B.K.

    1998-02-01

    The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

  19. Cylindrical reinforced-concrete pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Vaessen, F.

    1975-01-01

    The cylindrical pressure vessel has got a wall and an isolating layer composed of blocks of heat-resistant concrete or of ceramic material. The side of the isolating layer facing the interior of the presssure vessel is coated by a liner made of metallic material. In cold state and without internal pressure, the radius of this liner is smaller by a differential amount than that of the isolating layer. By means of radially displaceable fixing elements consisting of an anchoring tube and a holding tube inserted in it, the liner can be made to rest against the isolating layer. This occurs if the pressure vessel is brought to operational temperature. The anchoring tube is attached to the isolating layer whereas the displaceable holding tube is connected with the liner. The possible relative travelling distance of these two elements is equal to the difference of length of the two radii. In addition, the liner may consist of single parts connected with each other through compensating flanges. There may also be additional springs arranged between the isolating layer and the liner. (DG/PB) [de

  20. Composite Overwrapped Pressure Vessels (COPV) Stress Rupture Test

    Science.gov (United States)

    Russell, Richard; Flynn, Howard; Forth, Scott; Greene, Nathanael; Kezian, Michael; Varanauski, Don; Yoder, Tommy; Woodworth, Warren

    2009-01-01

    One of the major concerns for the aging Space Shuttle fleet is the stress rupture life of composite overwrapped pressure vessels (COPVs). Stress rupture life of a COPV has been defined as the minimum time during which the composite maintains structural integrity considering the combined effects of stress levels and time. To assist in the evaluation of the aging COPVs in the Orbiter fleet an analytical reliability model was developed. The actual data used to construct this model was from testing of COPVs constructed of similar, but not exactly same materials and pressure cycles as used on Orbiter vessels. Since no actual Orbiter COPV stress rupture data exists the Space Shuttle Program decided to run a stress rupture test to compare to model predictions. Due to availability of spares, the testing was unfortunately limited to one 40" vessel. The stress rupture test was performed at maximum operating pressure at an elevated temperature to accelerate aging. The test was performed in two phases. The first phase, 130 F, a moderately accelerated test designed to achieve the midpoint of the model predicted point reliability. The more aggressive second phase, performed at 160 F was designed to determine if the test article will exceed the 95% confidence interval of the model. This paper will discuss the results of this test, it's implications and possible follow-on testing.

  1. Prestressed concrete pressure vessel for a nuclear reactor

    International Nuclear Information System (INIS)

    Ritz, L.

    1976-01-01

    A prestressed concrete pressure vessel is described which is to be used for a nuclear reactor. It is of integrated construction, has a fixed amount of prestressing and can be used at very high internal pressures. The re-inforcement is arranged in U shapes in the concrete and the space of the reactor core and equipment is situated between the two uprights (of the U). Their ends are anchored to the outer walls of the concrete. All ends of the cross braced re-inforcement are equally distributed over the circumference of the common concrete structure along its height. (UWI) [de

  2. Composite Pressure Vessel Variability in Geometry and Filament Winding Model

    Science.gov (United States)

    Green, Steven J.; Greene, Nathanael J.

    2012-01-01

    Composite pressure vessels (CPVs) are used in a variety of applications ranging from carbon dioxide canisters for paintball guns to life support and pressurant storage on the International Space Station. With widespread use, it is important to be able to evaluate the effect of variability on structural performance. Data analysis was completed on CPVs to determine the amount of variation that occurs among the same type of CPV, and a filament winding routine was developed to facilitate study of the effect of manufacturing variation on structural response.

  3. Prevention of catastrophic failure in pressure vessels and pipings

    International Nuclear Information System (INIS)

    Rintamaa, R.; Wallin, K.; Ikonen, K.; Toerroenen, K.; Talja, H.; Keinaenen, H.; Saarenheimo, A.; Nilsson, F.; Sarkimo, M.; Waestberg, S.; Debel, C.

    1989-01-01

    The fracture resistance and integrity of pressure-loaded components have been assessed in a Nordic research programme. Experiments were performed to validate the computational fracture assessment analysis. Two tests were also conducted on a large decommissioned pressure vessel from an oil refinery plant. Different fracture assessment methods were developed and subsequently applied to the tested components. Interlaboratory round robin programmes with the participation of several laboratories were arranged to examine elastic-plastic finit element calculations and fracture mechanics testing. The transferability of material parameters derived from small specimens with simple crack geometries to more realistic crack geometries in real components has been verified. (author)

  4. Terahertz NDE of Stressed Composite Overwrapped Pressure Vessels - Initial Testing

    Science.gov (United States)

    Madaras, Eric I.; Seebo, Jeffrey P.; Anatasi, Robert F.

    2009-01-01

    Terahertz radiation nondestructive evaluation was applied to a set of Kevlar composite overwrapped pressure vessel bottles that had undergone a series of thermal and pressure tests to simulate stress rupture effects. The bottles in these nondestructive evaluation tests were bottles that had not ruptured but had survived various times at the elevated load and temperature levels. Some of the bottles showed evidence of minor composite failures. The terahertz radiation did detect visible surface flaws, but did not detect any internal chemical or material degradation of the thin overwraps.

  5. Dual shell pressure balanced reactor vessel. Final project report

    International Nuclear Information System (INIS)

    Robertus, R.J.; Fassbender, A.G.

    1994-10-01

    The Department of Energy's Office of Energy Research (OER) has previously provided support for the development of several chemical processes, including supercritical water oxidation, liquefaction, and aqueous hazardous waste destruction, where chemical and phase transformations are conducted at high pressure and temperature. These and many other commercial processes require a pressure vessel capable of operating in a corrosive environment where safety and economy are important requirements. Pacific Northwest Laboratory (PNL) engineers have recently developed and patented (U.S. patent 5,167,930 December 1, 1992) a concept for a novel Dual Shell Pressure Balanced Vessel (DSPBV) which could solve a number of these problems. The technology could be immediately useful in continuing commercialization of an R ampersand D 100 award-winning technology, Sludge-to-oil Reactor System (STORS), originally developed through funding by OER. Innotek Corporation is a small business that would be one logical end-user of the DSPBV reactor technology. Innotek is working with several major U.S. engineering firms to evaluate the potential of this technology in the disposal of wastes from sewage treatment plants. PNL entered into a CRADA with Innotek to build a bench-scale demonstration reactor and test the system to advance the economic feasibility of a variety of high pressure chemical processes. Hydrothermal processing of corrosive substances on a large scale can now be made significantly safer and more economical through use of the DSPBV. Hydrothermal chemical reactions such as wet-air oxidation and supercritical water oxidation occur in a highly corrosive environment inside a pressure vessel. Average corrosion rates from 23 to 80 miles per year have been reported by Rice (1994) and Latanision (1993)

  6. 46 CFR 154.650 - Cargo tank and process pressure vessel welding.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo tank and process pressure vessel welding. 154.650... Equipment Construction § 154.650 Cargo tank and process pressure vessel welding. (a) Cargo tank and process pressure vessel welding must meet Subpart 54.05 and Part 57 of this chapter. (b) Welding consumables used...

  7. 46 CFR 167.25-5 - Inspection of boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Inspection of boilers, pressure vessels, piping and...) NAUTICAL SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-5 Inspection of boilers, pressure vessels, piping and appurtenances. The inspection of boilers, pressure vessels, piping and appurtenances...

  8. 46 CFR 50.05-5 - Existing boilers, pressure vessels or piping systems.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Existing boilers, pressure vessels or piping systems. 50... ENGINEERING GENERAL PROVISIONS Application § 50.05-5 Existing boilers, pressure vessels or piping systems. (a) Whenever doubt exists as to the safety of an existing boiler, pressure vessel, or piping system, the marine...

  9. 46 CFR 167.25-1 - Boilers, pressure vessels, piping and appurtenances.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Boilers, pressure vessels, piping and appurtenances. 167... SCHOOLS PUBLIC NAUTICAL SCHOOL SHIPS Marine Engineering § 167.25-1 Boilers, pressure vessels, piping and... the following standards for boilers, pressure vessels, piping and appurtenances: (1) Marine...

  10. 10 CFR 50.66 - Requirements for thermal annealing of the reactor pressure vessel.

    Science.gov (United States)

    2010-01-01

    ... Requirements for thermal annealing of the reactor pressure vessel. (a) For those light water nuclear power... operating plan must include: (i) A detailed description of the pressure vessel and all structures and... For those cases where materials are removed from the beltline of the pressure vessel, the stress...

  11. 46 CFR 78.33-1 - Repairs of boiler and pressure vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 3 2010-10-01 2010-10-01 false Repairs of boiler and pressure vessels. 78.33-1 Section... OPERATIONS Reports of Accidents, Repairs, and Unsafe Equipment § 78.33-1 Repairs of boiler and pressure vessels. (a) Before making any repairs to boilers or unfired pressure vessels, the chief engineer shall...

  12. 77 FR 59408 - Finding of Equivalence; Alternate Pressure Relief Valve Settings on Certain Vessels Carrying...

    Science.gov (United States)

    2012-09-27

    ... Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section VIII with respect to...; Alternate Pressure Relief Valve Settings on Certain Vessels Carrying Liquefied Gases in Bulk AGENCY: Coast...-ENG Policy Letter 04-12, ``Alternative Pressure Relief Valve Settings on Vessels Carrying Liquefied...

  13. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  14. Calculation of reactor pressure vessel fluence using TORT code

    International Nuclear Information System (INIS)

    Shin, Chul Ho; Kim, Jong Kyung

    1998-01-01

    TORT is employed for fast neutron fluence calculation at the reactor pressure vessel. KORI Unit 1 reactor at cycle 1 is modeled for this calculation. Three-dimensional cycle averaged assembly power distributions for KORI Unit 1 at cycle 1 are calculated by using the core physics code, NESTLE 5.0. The root mean square error is within 4.3% compared with NDR (Nuclear Design Report) for all burnup steps. The C/E (Calculated/Experimental) values for the in-vessel dosimeters distribute between 0.98 and 1.36. The most updated cross-section library, BUGLE-96 based on ENDF/B-VI is used for the neutron fluence calculation. The maximum fast neutron fluence calculated on reactor pressure vessel for KORI Unit 1 operated for 411.41 effective full power days is 1.784x10 18 n/cm 2 . The position of the maximum neutron fluence in RPV wall 1/4 T is nearby 60 cm below the midplane at zero degree

  15. Effect of defects induced by doping and fast neutron irradiation on the thermal properties of lithium ammonium sulphate crystals

    Energy Technology Data Exchange (ETDEWEB)

    Kandil, S.H.; Ramadan, T.A.; Darwish, M.M. (Alexandria Univ. (Egypt). Dept. of Materials Science); Kassem, M.E.; El-Khatib, A.M. (Alexandria Univ. (Egypt). Dept. of Physics)

    1994-05-01

    Structural defects were introduced in lithium ammonium sulphate crystals (LAS) either in the process of crystal growth (in the form of foreign ions) or by neutron irradiation. The effect of such defects on the thermal properties of LAS crystals was studied in the temperature range 300-500 K. It was assumed that the doped LAS crystals are composed of a two-phase system having different thermal parameters in each phase. The specific heat at constant pressure, C[sub p], of irradiated samples was found to decrease with increasing irradiation doses. The thermal expansion of LAS crystals was found to be dependent on neutron irradiation, and was attributed to two processes: the release of new species and the trapping process. (author).

  16. Study on neutron irradiation behavior of beryllium as neutron multiplier

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  17. Study on neutron irradiation behavior of beryllium as neutron multiplier

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo

    1998-01-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  18. In vitro antileishmanial properties of neutron-irradiated meglumine antimoniate

    Energy Technology Data Exchange (ETDEWEB)

    Borborema, Samanta Etel Treiger; Nascimento, Nanci do [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), SP (Brazil). Lab. de Biologia Molecular]. E-mail: samanta@usp.br; Osso Junior, Joao Alberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), SP (Brazil). Centro de Radiofarmacia]. E-mail: jaosso@ipen.br; Andrade Junior, Heitor Franco de [Instituto de Medicina Tropical de Sao Paulo (IMT-SP), SP (Brazil). Lab. de Protozoologia]. E-mail:hfandrad@usp.br

    2005-10-15

    Pentavalent antimony, as meglumine antimoniate (Glucantime) or sodium stibogluconate (Pentostam), is the main treatment for leishmaniasis, a complex of diseases caused by the protozoan Leishmania, and an endemic and neglected threat in Brazil. Despite over half a century of clinical use, their mechanism of action, toxicity and pharmacokinetic data remain unknown. The analytical methods for determination of antimony in biological systems remain complex and have low sensitivity. Radiotracer studies have a potential in pharmaceutical development. The aim of this study was to obtain a radiotracer for antimony, with suitable physical and biological properties. Meglumine antimoniate was neutron irradiated inside the IEA-R1 nuclear reactor, producing two radioisotopes {sup 122} Sb and {sup 124} Sb, with high radionuclidic purity and good specific activity. This compound showed the same antileishmanial activity as the native compound. The use of the radiotracers, easily created by neutron irradiation, could be an interesting tool to solve important questions in antimonial pharmacology. (author)

  19. Separation of Protactinium from Neutron Irradiated Thorium Oxide

    International Nuclear Information System (INIS)

    Dominguez, G.; Gutierrez, L.; Ropero, M.

    1983-01-01

    The chemical separation of thorium and protactinium can be carried out by leaching most of the last one, about 95%, with aqueous HF from neutron irradiated thorium oxide. This leaching reaction la highly favored by the transformation reaction of the ThO 2 material into ThF 4 . For both reactions, leaching and transformation, the reagents concentration, agitation speed and temperature influences were studied and the activation energies were found. (Author) 18 refs

  20. National Low-Temperature Neutron-Irradiation Facility

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.; Klabunde, C.E.; Young, F.W. Jr.

    1983-08-01

    The Materials Sciences Division of the United States Department of Energy will establish a National Low Temperature Neutron Irradiation Facility (NLTNIF) which will utilize the Bulk Shielding Reactor (BSR) located at Oak Ridge National Laboratory. The facility will provide high radiation intensities and special environmental and testing conditions for qualified experiments at no cost to users. This report describes the planned experimental capabilities of the new facility

  1. Studies of neutron irradiation effects at IPNS-REF

    International Nuclear Information System (INIS)

    Kirk, M.A.

    1983-09-01

    Neutron irradiation effects studies at the Radiation Effects Facility (REF) at the Intense Pulsed Neutron Source (IPNS) located at Argonne National Laboratory (ANL) are reviewed. A brief history of the development of this user facility is followed by an overview of the scientific program. Experiments unique to a spallation neutron source are covered in more detail. Future direction of research at this facility is suggested

  2. Neutron irradiation effects on high Nicalon silicon carbide fibers

    Energy Technology Data Exchange (ETDEWEB)

    Osborne, M.C.; Steiner, D.; Snead, L.L. [Oak Ridge National Laboratory, TN (United States)

    1996-10-01

    The effects of neutron irradiation on the mechanical properties and microstructure of SiC and SiC-based fibers is a current focal point for the development of radiation damage resistant SiC/SiC composites. This report discusses the radiation effects on the Nippon Carbon Hi-Nicalon{trademark} fiber system and also discusses an erratum on earlier results published by the authors on this material. The radiation matrix currently under study is also summarized.

  3. Neutron irradiation effects on high Nicalon silicon carbide fibers

    International Nuclear Information System (INIS)

    Osborne, M.C.; Steiner, D.; Snead, L.L.

    1996-01-01

    The effects of neutron irradiation on the mechanical properties and microstructure of SiC and SiC-based fibers is a current focal point for the development of radiation damage resistant SiC/SiC composites. This report discusses the radiation effects on the Nippon Carbon Hi-Nicalon trademark fiber system and also discusses an erratum on earlier results published by the authors on this material. The radiation matrix currently under study is also summarized

  4. Evaluation of gamma and neutron irradiation effects on the ...

    Indian Academy of Sciences (India)

    However, appreciable change in the capacitance has been observed due to low doses of fast neutrons (cumulative dose, 115 cGy) with flux ∼ 9.925 × 107 neutrons/cm2 h from 252Cf neutron source of fluence, 2.5 × 107 neutrons/s. We have also observed that the impact of gamma and neutron irradiation is more at ...

  5. Characterization of the National Low-Temperature Neutron Irradiation Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kerchner, H.R.; Coltman, R.R. Jr.; Klabunde, C.E.; Young, F.W. Jr.

    1986-02-01

    The National Low-Temperature Neutron Irradiation Facility (NLTNIF) is now operating at the Bulk Shielding Reactor at ORNL. The facility provides high radiation intensities and special environmental and testing conditions for qualified experiments at no cost to users. A general description and major specifications of the NLTNIF are presented along with the results of performance tests. In addition, the hardware and other considerations required to perform experiments in the NLTNIF are described.

  6. Supplementary surveillance programme for WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Novosad, P.

    1994-01-01

    Paper gives a short analysis of a standard surveillance program for WWER440/213 type reactor pressure vessels. The program suggests the surveillance specimens are located in the region of high lead factor and thus they are suitable for surveillance for only up to about 5 years. For better and effective use of the results, their re-evaluation, based on gamma scanning of irradiated specimens as well as on testing of reconstituted irradiated specimens with the goal to determine the static and dynamic fracture toughness has been opened. New, supplementary surveillance program for the vessels operated in Czech Republic was designed: archive pieces of specimens (for further reconstitution after irradiation) will be irradiated in sites with low lead factor and then tested to determine the standard (impact) and fracture toughnesses

  7. Structural properties and neutron irradiation effects of ceramics

    International Nuclear Information System (INIS)

    Yano, Toyohiko

    1994-01-01

    In high temperature gas-cooled reactors and nuclear fusion reactors being developed at present, various ceramics are to be used in the environment of neutron irradiation for undertaking important functions. The change of the characteristics of those materials by neutron irradiation must be exactly forecast, but it has been known that the response of the materials is different respectively. The production method of ceramics and the resulted structures of ceramics which control their characteristics are explained. The features of covalent bond and ionic bond, the synthesis of powder and the phase change by heating, sintering and sintering agent, and grain boundary phase are described. The smelling of ceramics by neutron irradiation is caused by the formation of the clusters of Frenkel defects and minute spot defects. Its restoration by annealing is explained. The defects remaining in materials after irradiation are the physical defects by flipping atoms cut due to the collision with high energy particles and the chemical defects by nuclear transformation. Some physical defects can be restored, but chemical defects are never restored. The mechanical properties of ceramics and the effect of irradiation on them, and the thermal properties of ceramics and the effect of irradiation on them are reported. (K.I.)

  8. Evaluation of Data-Logging Transducer to Passively Collect Pressure Vessel p/T History

    Science.gov (United States)

    Wnuk, Stephen P.; Le, Son; Loew, Raymond A.

    2013-01-01

    Pressure vessels owned and operated by NASA are required to be regularly certified per agency policy. Certification requires an assessment of damage mechanisms and an estimation of vessel remaining life. Since detail service histories are not typically available for most pressure vessels, a conservative estimate of vessel pressure/temperature excursions is typically used in assessing fatigue life. This paper details trial use of a data-logging transducer to passively obtain actual pressure and temperature service histories of pressure vessels. The approach was found to have some potential for cost savings and other benefits in certain cases.

  9. Treating asphericity in fuel particle pressure vessel modeling

    Science.gov (United States)

    Miller, Gregory K.; Wadsworth, Derek C.

    1994-07-01

    The prototypical nuclear fuel of the New Production Modular High Temperature Gas-Cooled Reactor (NP-MHTGR) consists of spherical TRISO-coated particles suspended in graphite cylinders. The coating layers surrounding the fuel kernels in these particles consist of pyrolytic carbon layers and a silicon carbide layer. These coating layers act as a pressure vessel which retains fission product gases. In the operating conditions of the NP-MHTGR, a small percentage of these particles (pressure vessels) are expected to fail due to the pressure loading. The fuel particles of the NP-MHTGR deviate to some degree from a true spherical shape, which may have some effect on the failure percentages. A method is presented that treats the asphericity of the particles in predicting failure probabilities for particle samples. It utilizes a combination of finite element analysis and Monte Carlo sampling and is based on the Weibull statistical theory. The method is used here to assess the effects of asphericity in particles of two common geometric shapes, i.e. faceted particles and ellipsoidal particles. The method presented could be used to treat particle anomalies other than asphericity.

  10. Ultrasonic examination of austenitic welds at reactor pressure vessels

    International Nuclear Information System (INIS)

    Edelmann, X.

    1991-01-01

    Austenitic welds play an important role in reactor pressure vessels (RPV) both of pressurized water reactors (PWRs) and boiling water reactors (BWRs). These pressure retaining joints still pose problems for ultrasonic examination. During the last few years some progress has been achieved in overcoming these problems. The state of the art of the practical application is described in 'The Handbook on Ultrasonic Examination of Austenitic Welds' of the International Institute of Welding which was issued in 1986. This document was prepared by an international working group. Sulzer Brothers Limited, Winterthur Switzerland as manufacturer of both BWR and PWR RPVs has a long experience with ultrasonic examination of austenitic welds. In General Electric BWR RPVs the main recirculation line, the feedwater line and additional lines are connected to nozzles of the RPV with dissimilar metal welds joining stainless steel to carbon steel. The control rod drive (CRD) housings are attached to the RPV with complex weld joints. Framatome and Westinghouse RPVs also are provided with welds between the vessel and the main coolant line. This paper presents examples of ultrasonic approaches and examination results. Aspects of fabrication, preservice and inservice inspections are also covered. Complex weld joint design both from the point of material and geometry make ultrasonic examination sometimes extremely difficult but based on case by case investigations reliable solutions can often be realized. (orig.)

  11. Inspection device for external examination of pressure vessels, preferably for ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Figlhuber, D.; Gallwas, J.; Weber, R.; Weber, J.

    1978-01-01

    The inspection device is placed in the annular gap between pressure vessel and biological shield of the BWR. In the annulus there is arranged at least one longitudinal rail which has got vertical guideways. Along it there can be moved on testing paths a manipulator with the ultrasonic search unit. The manipulator drive is outside of the inspection annulus. It is coupled to the manipulator by means of a tension member being guided over a reversing unit mounted at the upper end of the longitudinal rail. As a tension member there may be used a drag chain; the drive and the reversing unit are provided with corresponding chain wheels. (DG) [de

  12. Study of radiation hardening in reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Nogiwa, Kimihiro; Nishimura, Akihiko

    2008-01-01

    In order to investigate the dependence of hardening on copper precipitate diameter and density, in-situ transmission electron microscopy (TEM) observations during tensile tests of dislocation gliding through copper rich-precipitates in thermally aged and neutron irradiated Fe-Cu alloys were performed. The obstacle strength has been estimated from the critical bow-out angle, φ, of dislocations. The obstacle distance on the dislocation line measured from in-situ TEM observations were compared with number density and diameter measured by 3D-AP (three dimensional atom probe) and TEM observation. A comparison is made between hardening estimation based on the critical bowing angles and those obtained from conventional tensile tests. (author)

  13. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    1996-01-01

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  14. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

    2017-11-15

    One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

  15. Corrosion fatigue of pressure vessel steel A533B. Part

    International Nuclear Information System (INIS)

    Gott, K.

    1978-08-01

    A survey is presented of the current work, both reported and in progress, to study the corrosion fatigue of the pressure vessel steel A533B C1 1 in a BWR environment. This report is based on the available literature, laboratory visits, and attendance at the Metals Society conference (April 1978). The parameters which are considered to have the most influence on the fatigue crack growth rate are considered in turn. Because of the sparsity of the available information an International Cooperative Group on Cyclic Crack Growth Rate Testing and Evaluation has been established. The initial work of the group and Swedish participation are described. (author)

  16. Composite Overwrapped Pressure Vessels (COPV) Materials Aging Issues

    Science.gov (United States)

    2010-01-01

    This slide presentation reviews some of the issues concerning the aging of the materials in a Composite Overwrapped Pressure Vessels (COPV). The basic composition of the COPV is a Boss, a composite overwrap, and a metallic liner. The lifetime of a COPV is affected by the age of the overwrap, the cyclic fatigue of the metallic liner, and stress rupture life, a sudden and catastrophic failure of the overwrap while holding at a stress level below the ultimate strength for an extended time. There is information about the coupon tests that were performed, and a test on a flight COPV.

  17. Reactor pressure vessel embrittlement: Insights from neural network modelling

    Science.gov (United States)

    Mathew, J.; Parfitt, D.; Wilford, K.; Riddle, N.; Alamaniotis, M.; Chroneos, A.; Fitzpatrick, M. E.

    2018-04-01

    Irradiation embrittlement of steel pressure vessels is an important consideration for the operation of current and future light water nuclear reactors. In this study we employ an ensemble of artificial neural networks in order to provide predictions of the embrittlement using two literature datasets, one based on US surveillance data and the second from the IVAR experiment. We use these networks to examine trends with input variables and to assess various literature models including compositional effects and the role of flux and temperature. Overall, the networks agree with the existing literature models and we comment on their more general use in predicting irradiation embrittlement.

  18. Contributions of Cu-rich clusters, dislocation loops and nanovoids to the irradiation-induced hardening of Cu-bearing low-Ni reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Bergner, F., E-mail: f.bergner@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany); Gillemot, F. [Centre for Energy Research of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Hernández-Mayoral, M.; Serrano, M. [Division of Materials, CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Török, G. [Wigner Research Center for Physics of the Hungarian Academy of Sciences, 29-33 Konkoly-Thege street, 1121 Budapest XII (Hungary); Ulbricht, A.; Altstadt, E. [Helmholtz-Zentrum Dresden-Rossendorf, Bautzner Landstr. 400, 01328 Dresden (Germany)

    2015-06-15

    Highlights: • TEM and SANS were applied to estimate mean size and number density of loops, nanovoids and Cu-rich clusters. • A three-feature dispersed-barrier hardening model was applied to estimate the yield stress increase. • The values and errors of the dimensionless obstacle strength were estimated in a consistent way. • Nanovoids are stronger obstacles for dislocation glide than dislocation loops, loops are stronger than Cu-rich clusters. • For reactor-relevant conditions, Cu-rich clusters contribute most to hardening due to their high number density. - Abstract: Dislocation loops, nanovoids and Cu-rich clusters (CRPs) are known to represent obstacles for dislocation glide in neutron-irradiated reactor pressure vessel (RPV) steels, but a consistent experimental determination of the respective obstacle strengths is still missing. A set of Cu-bearing low-Ni RPV steels and model alloys was characterized by means of SANS and TEM in order to specify mean size and number density of loops, nanovoids and CRPs. The obstacle strengths of these families were estimated by solving an over-determined set of linear equations. We have found that nanovoids are stronger than loops and loops are stronger than CRPs. Nevertheless, CRPs contribute most to irradiation hardening because of their high number density. Nanovoids were only observed for neutron fluences beyond typical end-of-life conditions of RPVs. The estimates of the obstacle strength are critically compared with reported literature data.

  19. Test of 6-inch-thick pressure vessels. Series 2. Intermediate test vessels V-3, V-4, and V-6

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Raftenberg, M.N.; Robinson, G.C.; Smith, J.E.

    1975-11-01

    The second series of intermediate vessel tests were crack initiation fracture tests of 6-in.-thick 39-in.-OD steel vessels with sharp surface flaws approximately 2 1 / 2 in. deep by 8 in. long in the longitudinal weld seams of the test cylinders. Fracture was initiated by means of hydraulic pressurization. One vessel was tested at each of three temperatures: 75, 130, and 190 0 F. Pretest analyses were made to predict the failure pressures and strains. Fracture toughness data obtained by equivalent-energy analysis of precracked Charpy-V tests and compact-tension specimen tests were used in the fracture analyses. The vessels behaved generally as had been expected. Posttest fracture analyses were also performed for each vessel. Detailed discussions of the fracture analysis methods developed in support of the vessel tests described are included. 34 references

  20. Ver reactor pressure vessel neutron exposure evaluation and monitoring

    International Nuclear Information System (INIS)

    Osmera, B.

    1992-01-01

    The surveillance neutron exposure monitoring system In VVER-440, V 213 series, consists of 54 Fe(n,p), 63 Cu(n,α and 93 Nb(n,n) detectors. The Fe detector is loaded for one year Cone fuel cycle) Irradiation only. The evaluation of the neutron monitor results including transformation to the critical locations of the reactor pressure vessel (RPV) can be based on the direct estimations of the neutron spectra and led factors (and other important space energy indices) in the mock-up experiments carried out. in the LR-0 experimental reactor, N.R.I. Rez. Because of high acceleration factor (about 11 for neutrons with energy above 0.5 MeV) the specimens are pulled out within five years of VVER-440 operation. After that the ex-vessel monitoring should be applied for RPV exposure determination. A recommended procedure (guide) for the RPV neutron exposure evaluation has been completed recently. All previous experimental data (the surveillance and ex-vessel monitors, I.R-0 mock-up measurements) have been revised arid reevaluated using the IRDF-90 (IAEA) cross section data. A brief review of this work is presented in the paper. Some comments relating to the RPV dosimetry of VVER-440, series 230, and VVER-1000 reactors are also done. (author)

  1. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  2. Burst pressure investigation of filament wound type IV composite pressure vessel

    Science.gov (United States)

    Farhood, Naseer H.; Karuppanan, Saravanan; Ya, H. H.; Baharom, Mohamad Ariff

    2017-12-01

    Currently, composite pressure vessels (PVs) are employed in many industries such as aerospace, transportations, medical etc. Basically, the use of PVs in automotive application as a compressed natural gas (CNG) storage cylinder has been growing rapidly. Burst failure due to the laminate failure is the most critical failure mechanism for composite pressure vessels. It is predominantly caused by excessive internal pressure due to an overfilling or an overheating. In order to reduce fabrication difficulties and increase the structural efficiency, researches and studies are conducted continuously towards the proper selection of vessel design parameters. Hence, this paper is focused on the prediction of first ply failure pressure for such vessels utilizing finite element simulation based on Tsai-Wu and maximum stress failure criterions. The effects of laminate stacking sequence and orientation angle on the burst pressure were investigated in this work for a constant layered thickness PV. Two types of winding design, A [90°2/∓θ16/90°2] and B [90°2/∓θ]ns with different orientations of helical winding reinforcement were analyzed for carbon/epoxy composite material. It was found that laminate A sustained a maximum burst pressure of 55 MPa for a sequence of [90°2/∓15°16/90°2] while the laminate B returned a maximum burst pressure of 45 MPa corresponding to a stacking sequence of [90°2/±15°/90°2/±15°/90°2/±15° ....] up to 20 layers for a constant vessel thickness. For verification, a comparison was done with the literature under similar conditions of analysis and good agreement was achieved with a maximum difference of 4% and 10% for symmetrical and unsymmetrical layout, respectively.

  3. Stressing arrangement for pressure vessels with prestressed jacket

    International Nuclear Information System (INIS)

    1976-01-01

    Previously used stressing arrangements for pressure vessels assembled from segments are improved by producing defined prestressing forces. The manufactured elements made, for example, of cast iron, which make up the containment wall, carry one (or more) of the prestressing arrangements described in detail on the outside. The stressing arrangements of one position of the wall elements lie in the same horizontal plane around the container. A stressing cable or rope without or with only small prestressing is positioned around the prestressing arrangements lying at the same height. By operating the stressing piston arrangement of each device a required pressure is exerted on each wall element, and is permanently fixed by tightening self locking wedges or screws. (ORU) [de

  4. Needs for evaluated covariance data for reactor pressure vessel dosimetry

    International Nuclear Information System (INIS)

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

    1992-01-01

    This report discusses new methodology for quantifying and then reducing uncertainties in the calculated pressure vessel fluences of a pressurized water reactor (PWR). The technique involves combining the integral results of the calculated and measured PWR surveillance dosimetry activities with the differential data used in the calculations, along with covariances of all the quantities, into a generalized linear least-squares adjustment procedure. Based on analysis of both PWRs and test reactor benchmarks, substantial evidence now exists to support the conclusion that, of all the nuclear as well as non-nuclear differential data considered, ENDF/B-VI values of the total inelastic iron cross sections and their covariances are the most important data controlling the outcome of the adjustment procedure. Predicted adjustments in these cross sections provided the stimulus for new measurements, the results of which impacted the ENDF/B-VI evaluation of iron 56

  5. Improved fireman's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    King, H. A.; Morris, E. E.

    1973-01-01

    Prototype high pressure glass filament-wound, aluminum-lined pressurant vessels suitable for use in a fireman's compressed air breathing system were designed, fabricated, and acceptance tested in order to demonstrate the feasibility of producing such high performance, lightweight units. The 4000 psi tanks have a 60 standard cubic foot (SCF) air capacity, and have a 6.5 inch diamter, 19 inch length, 415 inch volume, weigh 13 pounds when empty, and contain 33 percent more air than the current 45 SCF (2250 psi) steel units. The current steel 60 SCF (3000 psi) tanks weigh approximately twice as much as the prototype when empty, and are 2 inches, or 10 percent shorter. The prototype units also have non-rusting aluminum interiors, which removes the hazard of corrosion, the need for internal coatings, and the possibility of rust particles clogging the breathing system.

  6. Inspecting nuclear pressure vessels: the conundrum of minimizing risk

    International Nuclear Information System (INIS)

    Oestberg, G.

    1992-01-01

    The probability of a sudden, massive release of radioactivity from a light-water nuclear reactor through a breach of the containment is assessed on the basis of statistical data which partly consist of subjective estimates. This breach refers to the existence of crack-like defects remaining after a non-destructive examination of the main pressure vessel surrounding the reactor core. Two studies have recently been made of such sources of information about the effectiveness of non-destructive examination of pressure vessels with respect to defects. The results of these studies indicate that the data used as input in the probabilistic calculations do not possess the reliability that might be assumed from the assessments. This type of failure should therefore no longer be considered a de minimis case. In the present review the overconfidence in the efficiency of non-destructive examination is discussed from psychological, sociological and political science points of view. It is concluded that ingrained professional assumptions and values seem to be the main reason for the trust in the technology of inspection. However, there are also psychological constraints that can be understood only in their social and political contexts. (author)

  7. Flux effect analysis in WWER-440 reactor pressure vessel steels

    Science.gov (United States)

    Kryukov, A.; Blagoeva, D.; Debarberis, L.

    2013-11-01

    The results of long term research programme concerning the determination of irradiation embrittlement dependence on fast neutron flux for WWER-440 reactor pressure vessel steels before and after annealing are presented in this paper. The study of flux effect was carried out on commercial WWER-440 steels which differ significantly in phosphorous (0.013-0.036 wt%) and copper (0.08-0.20 wt%) contents. All specimens were irradiated in surveillance channel positions under similar conditions at high ˜4 × 1012 сm-2 s-1 and low ˜6 × 1011 сm-2 s-1 fluxes (E > 0.5 MeV) at a temperature of 270 °С. The radiation embrittlement was evaluated by transition temperature shift on the basis of Charpy specimens test results. In case of low flux, the measured Tk shifts could be 25-50 °C bigger than the Tk shifts obtained from high flux data. A significant flux effect is observed in WWER-440 reactor pressure vessel steels with higher copper content (>0.13 wt%).

  8. Corrosion of steel tendons used in prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Griess, J.C.; Naus, D.J.

    The purpose of this investigation was to determine the corrosion behavior of a high strength steel (ASTM A416-74 grade 270), typical of those used as tensioning tendons in prestressed concrete pressure vessels, in several corrosive environments and to demonstrate the protection afforded by coating the steel with either of two commercial petroleum-base greases or Portland Cement grout. In addition, the few reported incidents of prestressing steel failures in concrete pressure vessels used for containment of nuclear reactors are reviewed. The susceptibility of the steel to stress corrosion cracking and hydrogen embrittlement and its general corrosion rate were determined in several salt solutions. Wires coated with the greases and grout were soaked for long periods in the same solutions and changes in their mechanical properties were subsequently determined. All three coatings appeared to give essentially complete protection but small flaws in the grease coatings were detrimental; flaws or cracks less than 1 mm wide in the grout were without effect

  9. Microstructure and embrittlement of VVER 440 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Hennion, A.

    1999-03-01

    27 VVER 440 pressurised water reactors operate in former Soviet Union and in Eastern Europe. The pressure vessel, is made of Cr-Mo-V steel. It contains a circumferential arc weld in front of the nuclear core. This weld undergoes a high neutron flux and contains large amounts of copper and phosphorus, elements well known for their embrittlement potency under irradiation. The embrittlement kinetic of the steel is accelerated, reducing the lifetime of the reactor. In order to get informations on the microstructure and mechanical properties of these steels, base metals, HAZ, and weld metals have been characterized. The high amount of phosphorus in weld metals promotes the reverse temper embrittlement that occurs during post-weld heat treatment. The radiation damage structure has been identified by small angle neutron scattering, atomic probe, and transmission electron microscopy. Nanometer-sized clusters of solute atoms, rich in copper with almost the same characteristics as in western pressure vessels steels, and an evolution of the size distribution of vanadium carbides, which are present on dislocation structure, are observed. These defects disappear during post-irradiation tempering. As in western steels, the embrittlement is due to both hardening and reduction of interphase cohesion. The radiation damage specificity of VVER steels arises from their high amount of phosphorus and from their significant density of fine vanadium carbides. (author)

  10. Strain ageing in welds of nuclear pressure vessels

    International Nuclear Information System (INIS)

    Otterberg, R.; Karlsson, C.

    1979-01-01

    Static and dynamic strain ageing have been investigated on submerged-arc welds and repair welds from plates of the pressure vessel steel A 533B. The results permit the determination of the worst strain ageing conditions existing in a nuclear pressure vessel. Static strain ageing was investigated by means of data from tension tests, hardness measurements and Charpy-V impact properties for prestrained and aged material for ageing temperatures from room temperature to 350 deg C and ageing times up to 1000h. Dynamic strain ageing was investigated by tensile tests up to 350 deg C at different strain rates. At the most static strain ageing was found to increase the impact transition temperature from -75 deg C in the as-received condition to -55 deg C after prestraining and ageing for the plate material, from -35 to -10 deg C for the submerged arc weld and from -90 to -40 deg C for the repair weld. Approximately 10 deg C of the deleterious effect is due to the effect of ageing for the two former materials whereas the corresponding figure for the repair weld amounts to 35 deg C. The dynamic strain ageing is strongest at very low strain rates at temperatures just below 300 deg C. The effect of strain ageing can be reduced by stress relief heat treatment or by other means decreasing the content of nitrogen in solution. (author)

  11. Minimum weight design of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Boes, R.

    1975-01-01

    A method of non-linear programming for the minimization of the volume of rotationally symmetric prestressed concrete reactor pressure vessels is presented. It is assumed that the inner shape, the loads and the degree of prestressing are prescribed, whereas the outer shape is to be detemined. Prestressing includes rotational and vertical tension. The objective function minimizes the weight of the PCRV. The constrained minimization problem is converted into an unconstrained problem by the addition of interior penalty functions to the objective function. The minimum is determined by the variable metric method (Davidson-Fletcher-Powell), using both values and derivatives of the modified objective function. The one-dimensional search is approximated by a method of Kund. Optimization variables are scaled. The method is applied to a pressure vessel like for THTR. It is found that the thickness of the cylindrical wall may be reduced considerably for the load cases considered in the optimization. The thickness of the cover is reduced slightly. The largest reduction in wall thickness occurs at the junction of wall and cover. (Auth.)

  12. State-of-the-art and prospets for designing and constraction of prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Anon.

    1985-01-01

    Short review of reports submitted to the symposium on pressure vessels, which was conducted in Calgary (Canada), has been presented. New tendencies of designing of prestressed concrete pressure vessels (PCPV) for nuclear for nuclear reactors are noted. Construction of hot vessel liner is studied. A conclusion is drawn on prospects of PCPV creation

  13. Behavior of radioisotope in liquid neutron irradiated Pb-17Li eutectic

    International Nuclear Information System (INIS)

    Tebus, V.N.; Aksenov, B.S.; Klabukov, U.G.

    1994-01-01

    Investigation of radioisotope 210 Po evaporation from liquid neutron irradiated Pb- 17 Li eutectic has been performed by Knudsen method. Equilibrium 210 Po vapor pressures at temperatures 250-700 degrees C were found about 3-4 orders of magnitude less than that for pure Po and were closed to equilibrium vapor pressures of Po-Pb compound. It was proposed Po forms stable Po-Pb compounds in eutectic at temperatures up to 750-800 degrees C. But disintegrates during long storage owing to self irradiation. It was determined Po aerosol transfer with radio gases takes place at the melting period. Contamination is happened also under irradiated eutectic storage at room temperature owing to aggregate recoil characteristic of Po

  14. NASA Requirements for Ground-Based Pressure Vessels and Pressurized Systems (PVS). Revision C

    Science.gov (United States)

    Greulich, Owen Rudolf

    2017-01-01

    The purpose of this document is to ensure the structural integrity of PVS through implementation of a minimum set of requirements for ground-based PVS in accordance with this document, NASA Policy Directive (NPD) 8710.5, NASA Safety Policy for Pressure Vessels and Pressurized Systems, NASA Procedural Requirements (NPR) 8715.3, NASA General Safety Program Requirements, applicable Federal Regulations, and national consensus codes and standards (NCS).

  15. Contribution for the improvement of pressurized thermal shock assessment methodologies in PWR pressure vessels

    International Nuclear Information System (INIS)

    Gomes, Paulo de Tarso Vida

    2005-01-01

    The structural integrity assessment of nuclear reactor pressure vessel, concerned to Pressurized Thermal Shock (PTS) accidents, became a necessity and has been investigated since the eighty's. The recognition of the importance of PTS assessment has led the international nuclear technology community to devote a considerable research effort directed to the complete integrity assessment process of the Reactor Pressure Vessels (VPR). Researchers in Europe, Japan and U.S.A. have concentrated efforts in the VPR structural and fracture analysis, conducting experiments to best understand how specific factors act on the behavior of discontinuities, under PTS loading conditions. The main goal of this work is to study de structural behavior of an 'in scale' PWR nuclear reactor pressure vessel model, containing actual discontinuities, under loading conditions generated by a pressurized thermal shock. To construct the pressure vessel model utilized in this research, the approach developed by Barroso (1995) and based on likelihood studies, related to thermal-hydraulic behavior during the PTS was employed. To achieve the objective of this research, a new methodology to generate cracks, with known geometry and localization in the vessel model wall was developed. Additionally, an hydraulic circuit, able to flood the vessel model, heated to 300 deg C, with 10 m 3 of water at 8 deg C, in 170 seconds, was built. Thermo-hydraulic calculations using RELAP5/M0D 3.2.2γ computational code were done, to estimate the temperature profiles during the cooling time. The resulting data subsidized the thermo-structural calculations that were accomplished using ANSYS 7.01 computational code, for both 2D and 3D models. So, the stress profiles obtained with these calculations were associated with fracture mechanics concepts, to assess the crack growth behavior in the VPR model wall. After the PTS test, the VPR model was submitted to destructive and non-destructive inspections. The results

  16. Characterization of neutron irradiated, low-resistivity silicon detectors

    International Nuclear Information System (INIS)

    Angarano, M.M.; Bilei, G.M.; Ciampolini, P.P.; Giorgi, M.; Mihul, A.; Militaru, O.; Passeri, D.; Scorzoni, A.

    2002-01-01

    A complete electrical characterization of silicon detectors fabricated using low- (≅1.5 kΩ cm) and high- (>5 kΩ cm) resistivity substrates has been carried out. Measurements have been performed before and after neutron irradiation at several different fluences, up to 3x10 14 n cm -2 (1 MeV eq.). Experimental results have been compared with CAD-based simulations. A good agreement has been found, thus validating the CAD model predictions. The adoption of low-resistivity devices appears to have some definite advantages in terms of depletion voltage, which in turn results in better interstrip capacitance and interstrip resistance

  17. Gamma and neutron irradiation tests on commercial IC op amps

    International Nuclear Information System (INIS)

    Kennedy, E.J.; Morris, A.C. Jr.; Su, D.K.

    1985-01-01

    Experimental results of gamma and neutron irradiation tests on 30 types of integrated-circuit operational amplifiers from 11 manufacturers are presented. All units were low-cost, commercial-grade devices. Op amps were evaluated for changes in offset voltage, input bias current, power supply current, open-loop gain, gain-bandwidth product, slew rate, power-supply and common-mode rejection ratios. Bipolar transistor op amps with resistive collector load resistors for the input stage indicated the best radiation hardness

  18. Microstructure and mechanical properties of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, E.; Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Terai, T.; Tanaka, S.

    1998-01-01

    Microstructure and mechanical properties of the neutron irradiated beryllium with total fast neutron fluences of 1.3 - 4.3 x 10{sup 21} n/cm{sup 2} (E>1 MeV) at 327 - 616degC were studied. Swelling increased by high irradiation temperature, high fluence, and by the small grain size and high impurity. Obvious decreasing of the fracture stress was observed in the bending test and in small grain specimens which had many helium bubbles on the grain boundary. Decreasing of the fracture stress for small grain specimens was presumably caused by crack propagation on the grain boundaries which weekend by helium bubbles. (author)

  19. Effects of neutron irradiation on a superconducting metallic glass

    International Nuclear Information System (INIS)

    Kramer, E.A.; Johnson, W.L.; Cline, C.

    1979-06-01

    The effects of fast neutron irradiation on a superconducting metallic glass (Mo 6 Ru 4 ) 82 B 18 have been studied. Following irradiation to a total fluence of 10 19 n/cm 2 , T/sub c/ increases from 6.05 K to 6.19 K, and the width of the transition decreases sharply. The density of the material decreases by 1.5%, and the x-ray scattering intensity maxima are broadened. An improvement in the ductility of the samples is observed which together with the other observations suggests the production of defects having atomic scale dimensions and characterized by excess volume

  20. Positron lifetime study of neutron-irradiated molybdenum

    International Nuclear Information System (INIS)

    Hinode, Kenji; Tanigawa, Shoichiro; Kumakura, Hiroaki; Doyama, Masao; Shiraishi, Kensuke.

    1978-01-01

    Annealing behavior of fast-neutron-irradiated molybdenum was studied by means of positron lifetime technique. It was found that Stage III annealing can be mainly identified as the vacancy migration process from the detailed analyses of data. The void growth after successive high temperature annealings was clearly detected through the changes of positron lifetime parameters. An attempt to analyse the size distribution of voids from positron lifetime spectra was presented, and discussions on the evaluation of void concentration from positron data are also given. (author)

  1. Neutron irradiation studies of avalanche photodiodes using californium-252

    Science.gov (United States)

    Reucroft, S.; Rusack, R.; Ruuska, D.; Swain, J.

    1997-02-01

    Californium-252 is a convenient and copious source of neutrons of energies around 1 MeV, and provides many advantages over reactors for neutron irradiation studies of detector components. We describe here an experimental setup at Oak Ridge National Laboratory which has been constructed to study the performance of avalanche photodiodes in neutron fluences up to 10 13 neutrons/cm 2, similar to what is expected in parts of the CMS detector at the LHC. An irradiation study of some avalanche photodiodes is discussed, followed by a brief summary of results obtained.

  2. Characterization of neutron irradiated, low-resistivity silicon detectors

    CERN Document Server

    Angarano, M M; Giorgi, M; Bilei, G M; Mihul, A; Militaru, O; Passeri, D; Scorzoni, A

    2000-01-01

    A complete electrical characterization of silicon detectors fabricated using low-( ~ 1.5 kOhm cm) and high-( > 5 kOhm cm) resistivity substrates has been carried out. Measurements have been performed before and after neutron irradiation at several different fluences, up to 3x10^14 n cm^-2 ( 1 MeV eq.). Experimental results have been compared with CAD based simulations. A good agreement has been found, thus validating the CAD model predictions. The adoption of low resistivity devices appears to have some definite advantages in terms of depletion voltage, which in turn results in better interstrip resistance.

  3. Three-Dimensional Digital Image Correlation of a Composite Overwrapped Pressure Vessel During Hydrostatic Pressure Tests

    Science.gov (United States)

    Revilock, Duane M., Jr.; Thesken, John C.; Schmidt, Timothy E.

    2007-01-01

    Ambient temperature hydrostatic pressurization tests were conducted on a composite overwrapped pressure vessel (COPV) to understand the fiber stresses in COPV components. Two three-dimensional digital image correlation systems with high speed cameras were used in the evaluation to provide full field displacement and strain data for each pressurization test. A few of the key findings will be discussed including how the principal strains provided better insight into system behavior than traditional gauges, a high localized strain that was measured where gages were not present and the challenges of measuring curved surfaces with the use of a 1.25 in. thick layered polycarbonate panel that protected the cameras.

  4. Marine reactor pressure vessels dumped in the Kara Sea

    International Nuclear Information System (INIS)

    Mount, M.E.

    1997-01-01

    Between 1965 and 1988, 16 marine reactors from seven Russian submarines and the icebreaker Lenin, each of which suffered some form of reactor accident, were dumped in a variety of containments, using a number of sealing methods, at five sites in the Kara Sea. All reactors were dumped at sites that varied in depth from 12 to 300 m and six contained their spent nuclear fuel (SNF). This paper examines the breakdown of the reactor pressure vessel (RPV) barriers due to corrosion, with specific emphasis on those RPVs containing SNF. Included are discussions of the structural aspects of the steam generating installations and their associated RPVs, a summary of the disposal operations, assumptions on corrosion rates of structural and filler materials, and an estimate of the structural integrity of the RPVs at the present time (1996) and in the year 2015

  5. Multipurpose Pressure Vessel Scanner and Photon Doppler Velocimetry

    Science.gov (United States)

    Ellis, Tayera

    2015-01-01

    Critical flight hardware typically undergoes a series of nondestructive evaluation methods to screen for defects before it is integrated into the flight system. Conventionally, pressure vessels have been inspected for flaws using a technique known as fluorescent dye penetrant, which is biased to inspector interpretation. An alternate method known as eddy current is automated and can detect small cracks better than dye penetrant. A new multipurpose pressure vessel scanner has been developed to perform internal and external eddy current scanning, laser profilometry, and thickness mapping on pressure vessels. Before this system can be implemented throughout industry, a probability of detection (POD) study needs to be performed to validate the system’s eddy current crack/flaw capabilities. The POD sample set will consist of 6 flight-like metal pressure vessel liners with defects of known size. Preparation for the POD includes sample set fabrication, system operation, procedure development, and eddy current settings optimization. For this, collaborating with subject matter experts was required. This technical paper details the preparation activities leading up to the POD study currently scheduled for winter 2015/2016. Once validated, this system will be a proven innovation for increasing the safety and reliability of necessary flight hardware.Additionally, testing of frangible joint requires Photon Doppler Velocimetry (PDV) and Digital Image Correlation instrumentation. There is often noise associated with PDV data, which necessitates a frequency modulation (FM) signal-to-noise pre-test. Generally, FM radio works by varying the carrier frequency and mixing it with a fixed frequency source, creating a beat frequency which is represented by audio frequency that can be heard between about 20 to 20,000 Hz. Similarly, PDV reflects a shifted frequency (a phenomenon known as the Doppler Effect) from a moving source and mixes it with a fixed source frequency, which results in

  6. Crystal Plasticity Model of Reactor Pressure Vessel Embrittlement in GRIZZLY

    Energy Technology Data Exchange (ETDEWEB)

    Chakraborty, Pritam [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Biner, Suleyman Bulent [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Zhang, Yongfeng [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Spencer, Benjamin Whiting [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2015-07-01

    The integrity of reactor pressure vessels (RPVs) is of utmost importance to ensure safe operation of nuclear reactors under extended lifetime. Microstructure-scale models at various length and time scales, coupled concurrently or through homogenization methods, can play a crucial role in understanding and quantifying irradiation-induced defect production, growth and their influence on mechanical behavior of RPV steels. A multi-scale approach, involving atomistic, meso- and engineering-scale models, is currently being pursued within the GRIZZLY project to understand and quantify irradiation-induced embrittlement of RPV steels. Within this framework, a dislocation-density based crystal plasticity model has been developed in GRIZZLY that captures the effect of irradiation-induced defects on the flow stress behavior and is presented in this report. The present formulation accounts for the interaction between self-interstitial loops and matrix dislocations. The model predictions have been validated with experiments and dislocation dynamics simulation.

  7. Fracture probability evaluation of a LWR pressure vessel

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  8. Computer system for International Reactor Pressure Vessel Materials Database support

    International Nuclear Information System (INIS)

    Arutyunjan, R.; Kabalevsky, S.; Kiselev, V.; Serov, A.

    1997-01-01

    This report presents description of the computer tools for support of International Reactor Pressure Vessel Materials Database developed at IAEA. Work was focused on raw, qualified, processed materials data, search, retrieval, analysis, presentation and export possibilities of data. Developed software has the following main functions: provides software tools for querying and search of any type of data in the database; provides the capability to update the existing information in the database; provides the capability to present and print selected data; provides the possibility of export on yearly basis the run-time IRPVMDB with raw, qualified and processed materials data to Database members; provides the capability to export any selected sets of raw, qualified, processed materials data

  9. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Connor, J.J.

    1975-01-01

    Numerical procedures for predicting the nonlinear behaviour of a prestressed concrete reactor vessel over its design life are discussed. The numerical models are constructed by combining three-dimensional isoparametric finite elements which simulate the concrete, thin shell elements which simulate steel liner plates and layers of reinforcement steel, and axial elements for discrete prestressing cables. Nonlinearity under compressive stress, multi-dimensional cracking, shrinkage, and stress/temperature induced creep of concrete are considered in addition to the elastic plastic behaviour of the liner and reinforcing steel. The analysis of an actual PCRV is described. Stress contours and cracking patterns in the region of cutouts corresponding to operational pressure and temperature loads are illustrated. The effects of creep, unloading,and creep recovery are then shown. Lastly, a strategy for assessing the performance over its design life is discussed. (Auth.)

  10. Preparation of the Shippingport reactor pressure vessel shipping package

    International Nuclear Information System (INIS)

    Yannitell, D.M.

    1988-01-01

    Shippingport Station Decommissioning Project is the removal and shipment the Reactor Pressure Vessel (PRV) and its associated Neutron Shield Tank (NST) to the government owned Hanford Reservation in Richland, Washington. Engineering studies considered the alternatives for removal and shipment of the RPV/NST. These included segmentation for subsequent truck shipments, and one-piece removal with barge or rail shipment. Although the analysis indicated that current technology could be utilized to accomplish either alternative, one-piece removal of the RPV was selected as the safest, most cost effective method. When compared to segmentation, it was estimated that one-piece removal would reduce the duration of the Project by 1 year, reduce cost by $4 M, and result in a savings of radiation exposure of 150 man-Rem. Rail transportation of an integral RPV/NST package is not feasible due to the physical size of the package. 5 refs., 1 fig

  11. Low Temperature Irradiation Embrittlement of Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    The embrittlement trend curve development project for HFIR reactor pressure vessel (RPV) steels was carried out with three major tasks. Which are (1) data collection to match that used in HFIR steel embrittlement trend published in 1994 Journal Nuclear Material by Remec et. al, (2) new embrittlement data of A212B steel that are not included in earlier HFIR RPV trend curve, and (3) the adjustment of nil-ductility-transition temperature (NDTT) shift data with the consideration of the irradiation temperature effect. An updated HFIR RPV steel embrittlement trend curve was developed, as described below. NDTT( C) = 23.85 log(x) + 203.3 log (x) + 434.7, with 2- uncertainty of 34.6 C, where parameter x is referred to total dpa. The developed update HFIR RPV embrittlement trend curve has higher embrittlement rate compared to that of the trend curve developed in 1994.

  12. Reactor Pressure Vessel P-T Limit Curve Round Robin

    Energy Technology Data Exchange (ETDEWEB)

    Jang, C.H.; Moon, H.R.; Jeong, I.S. [Korea Electric Power Research Institute, Taejon (Korea)

    2002-07-01

    This report is the summary of the analysis results for the P-T Limit Curve construction which have been subjected to the round robin analysis. The purpose of the round robin is to compare the procedure and method used in various organizations to construct P-T limit curve to prevent brittle fracture of reactor pressure vessel of nuclear power plants. Each Participant used its own approach to construct the P-T limit curve and submitted the results, By analyzing the results, the reference procedure for the P-T limit curve could be established. This report include the results of the comparison of the procedure and method used by the participants, and sensitivity study of the key parameters. (author) 23 refs, 88 figs, 17 tabs.

  13. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    GRIFFIN, PATRICK J.

    1999-01-01

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  14. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Berg, S.; Loeseth, S.; Holand, I.

    1977-01-01

    A computational model for circular symmetric reinforced concrete shell problems is described. The model is based on the Finite Element Method. Non-linear stress-strain constitutive relations are used for the concrete, the reinforcement and for the liner. The reinforcement layers may be of different steel qualities. Each layer may be given a specified prestressing. This can be done at the beginning of the computations or the specific reinforcement layer can be considered inactive until a specified level of loading is reached. Thus, the prestressing procedure may also be analyzed in detail. Bond-slip effects are not accounted for. However, no bond may be assumed for prestressing cables by inserting special reinforcement elements. Several models of prestressed concrete reactor pressure vessels which have been tested up to rupture have been analysed. Analytical (numerical) models for reinforced concrete are also discussed on a more general basis. (Auth.)

  15. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    Energy Technology Data Exchange (ETDEWEB)

    GRIFFIN, PATRICK J.

    1999-09-14

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation.

  16. 30 CFR 57.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... NONMETAL MINES Compressed Air and Boilers § 57.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with...

  17. 30 CFR 56.13001 - General requirements for boilers and pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false General requirements for boilers and pressure... MINES Compressed Air and Boilers § 56.13001 General requirements for boilers and pressure vessels. All boilers and pressure vessels shall be constructed, installed, and maintained in accordance with the...

  18. High dose effects in neutron irradiated face-centered cubic metals

    International Nuclear Information System (INIS)

    Garner, F.A.; Toloczko, M.B.

    1993-06-01

    During neutron irradiation, most face-centered cubic metals and alloys develop saturation or quasi-steady state microstructures. This, in turn, leads to saturation levels in mechanical properties and quasi-steady state rates of swelling and creep deformation. Swelling initially plays only a small role in determining these saturation states, but as swelling rises to higher levels, it exerts strong feedback on the microstructure and its response to environmental variables. The influence of swelling, either directly or indirectly via second order mechanisms, such as elemental segregation to void surfaces, eventually causes major changes, not only in irradiation creep and mechanical properties, but also on swelling itself. The feedback effects of swelling on irradiation creep are particularly complex and lead to problems in applying creep data derived from highly pressurized creep tubes to low stress situations, such as fuel pins in liquid metal reactors

  19. Support device in a pressure vessel, particularly a reactor pressure vessel, to support against horizontal forces. Abstuetzeinrichtung an einem Druckbehaelter, insbesondere einem Reaktordruckbehaelter, gegen Horizontalkraefte

    Energy Technology Data Exchange (ETDEWEB)

    Dorner, H.; Harand, E.; Scholz, M.; Scheler, R.

    1986-04-17

    In a support device on a reactor pressure vessel of a PWR, a guide leading downwards is provided in the vertical pellet cartridge on the floor side of the reactor pressure vessel, which is located in a fixed support pipe. The support pipe is supported on a concreted baseplate in the horizontal direction, and can be screwed to this. The support pipe has a plate-shaped support foot to support it on the baseplate. The reactor pressure vessel, normally resting on its main coolant duct in the plane normal to the axis, can be effectively supported against horizontal forces by this additional support, without preventing thermal movement in the longitudinal axis.

  20. Results of neutron irradiation of liquid lithium saturated with deuterium

    International Nuclear Information System (INIS)

    Tazhibayeva, Irina; Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna; Lyublinski, Igor; Vertkov, Alexey; Mazzitelli, Giuseppe

    2017-01-01

    Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(−144/RT). • The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10 −13 cm −2 s −1 . The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10 9 exp(-144/RT). The T 2 molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

  1. Tritium release from neutron irradiated lithium-aluminium oxides

    International Nuclear Information System (INIS)

    Guggi, D.; Ihle, H.R.; Kurz, U.

    1976-01-01

    The release of tritium from neutron irradiated Li 5 AlO 4 (low temperature phase α-Li 5 AlO 4 ; high temperature phase: β-Li 5 AlO 4 ) and Li 2 O powders was studied. Some measurements were also carried out with γ-LiAlO 2 (samples from fused materials) in order to investigate the effect of grain size on the release of tritium. The results on the tritium release from neutron irradiated powdered α- and β-Li 5 AlO 4 and Li 2 O expressed as time constant tau=r 2 /D.π 2 as a function of temperature are given by the following equations: α-Li 5 AlO 4 :ln(1/tau)=-(4460+-720)/T-(1.13+-0.14); β-Li 5 AlO 4 :ln(1/tau)=-(9000+-1200)/T+(6.06+-0.13); Li 2 O:ln(1/tau)=-(3460+-470)/T-(2.48+-0.12). From these equations it is seen that at elevated temperature, e.g. at 600 0 C tritium is released at a much higher rate from β-Li 5 AlO 4 than from α-Li 5 AlO 4 and Li 2 O

  2. Modelling of pressure increase protection system for the vacuum vessel of W7-X device

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Tadas, E-mail: tadas.kaliatka@lei.lt; Uspuras, Eugenijus; Kaliatka, Algirdas

    2016-11-01

    Highlights: • Two in-vessel LOCAs (partial and guillotine break of 40 mm diameter pipe of cooling system) for Wendelstein 7-X fusion device were analyzed. • The analysis of the processes in the cooling system, vacuum vessel and pressure increase protection system were performed using thermal-hydraulic RELAP5 Mod3.3 code. • The suitability of pressure increase protection system was assessed. - Abstract: In fusion devices, plasma is contained in a vacuum vessel. The vacuum vessel cannot withstand a pressure above atmospheric. Any damage of in-vessel components could lead to water ingress and may lead to pressure increase and possible damage of vacuum vessel. In order to avoid such undesirable consequences, the pressure increase protection system is designed. In this article, the processes occurring in the vacuum vessel and pressure increase protection system of W7-X device during LOCA (small and guillotine pipe break) event are analyzed. The model of W7-X cooling system, vacuum vessel and pressure increase protection system was developed using RELAP5 code. Numerical analysis of partial and guillotine break of 40 mm diameter pipe of cooling system was performed. Calculation results showed that burst disc of the pressure increase protection system does not open when the cross section area of partial break in the cooling system is smaller than 1 mm{sup 2}. During the guillotine break of cooling system, the burst disc opens, but pressure increase protection system is capable to prevent overpressure of the vacuum vessel.

  3. NDE and Stress Monitoring on Composite Overwrapped Pressure Vessels, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Damage caused by composite overwrapped pressure vessels (COPVs) failure can be catastrophic. Thus, monitoring condition and stress in the composite overwrap,...

  4. Cavity nucleation and growth during helium implantation and neutron irradiation of Fe and steel

    DEFF Research Database (Denmark)

    Eldrup, Morten Mostgaard; Singh, Bachu Narain

    In order to investigate the role of He in cavity nucleation in neutron irradiated iron and steel, pure iron and Eurofer-97 steel have been He implanted and neutron irradiated in a systematic way at different temperatures, to different He and neutron doses and with different He implantation rates....

  5. A mathematical model for pressure-based organs behaving as biological pressure vessels.

    Science.gov (United States)

    Casha, Aaron R; Camilleri, Liberato; Gauci, Marilyn; Gatt, Ruben; Sladden, David; Chetcuti, Stanley; Grima, Joseph N

    2018-04-26

    We introduce a mathematical model that describes the allometry of physical characteristics of hollow organs behaving as pressure vessels based on the physics of ideal pressure vessels. The model was validated by studying parameters such as body and organ mass, systolic and diastolic pressures, internal and external dimensions, pressurization energy and organ energy output measurements of pressure-based organs in a wide range of mammals and birds. Seven rules were derived that govern amongst others, lack of size efficiency on scaling to larger organ sizes, matching organ size in the same species, equal relative efficiency in pressurization energy across species and direct size matching between organ mass and mass of contents. The lung, heart and bladder follow these predicted theoretical relationships with a similar relative efficiency across various mammalian and avian species; an exception is cardiac output in mammals with a mass exceeding 10kg. This may limit massive body size in mammals, breaking Cope's rule that populations evolve to increase in body size over time. Such a limit was not found in large flightless birds exceeding 100kg, leading to speculation about unlimited dinosaur size should dinosaurs carry avian-like cardiac characteristics. Copyright © 2018. Published by Elsevier Ltd.

  6. Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

    International Nuclear Information System (INIS)

    Frid, W.

    1987-01-01

    This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals and gas bubble nucleation in molten metals are relevant problems which are addressed in this work. Models are developed for jet expansion, primary breakup of the jet and secondary fragmentation of melt droplets resulting from violent effervescence of dissolved gas. The jet expansion model is based on a general relation for bubble growth which includes both inertia-controlled and diffusion-controlled growth phases. The jet expansion model is able to predict the jet void fraction, jet radius as a function of axial distance from the pressure vessel, bubble size and bubble pressure. The number density of gas bubbles in the melt, which is a basic parameter in the model, was determined experimentally and is about 10 8 per m 3 of liquid. The primary breakup of the jet produces a spray of droplets, about 2-3 mm in diameter. Parametric calculations for a TMLB' reactor accident sequence show that the corium jet is disrupted within a few initial jet diameters from the reactor vessel and that the radius of corium spray at the level of the reactor cavity floor is in the range of 0.8 to 2.6 m. (orig./HP)

  7. Effects of shear forces and pressure on blood vessel function and metabolism in a perfusion bioreactor.

    Science.gov (United States)

    Hoenicka, Markus; Wiedemann, Ludwig; Puehler, Thomas; Hirt, Stephan; Birnbaum, Dietrich E; Schmid, Christof

    2010-12-01

    Bovine saphenous veins (BSV) were incubated in a perfusion bioreactor to study vessel wall metabolism and wall structure under tissue engineering conditions. Group 1 vessels were perfused for 4 or 8 days. The viscosity of the medium was increased to that of blood in group 2. Group 3 vessels were additionally strained with luminal pressure. Groups 1-d through 3-d were similar except that BSV were endothelium-denuded before perfusion. Groups 1-a through 3-a used native vessels at elevated flow rates. Group 3 vessels responded significantly better to noradrenaline on day 4, whereas denuded vessels showed attenuated responses (p vessels. pO₂ gradients across the vessels were independent of time and significantly higher in group 2 (p vessels of groups 3, 1-d, and 3-d which released more lactate than glucose could supply (p vessels as well as all vessels perfused with elevated flow rates showed a loss of endothelial cells after 4 days, whereas group 2 and 3 vessels retained most of the endothelium. These data suggest that vessel metabolism was not limited by oxygen supply. Shear forces did not affect glucose metabolism but increased oxygen consumption and endothelial cell survival. Luminal pressure caused the utilization of energy sources other than glucose, as long as the endothelium was intact. Therefore, vessel metabolism needs to be monitored during tissue engineering procedures which challenge the constructs with mechanical stimuli.

  8. Creep of A508/533 Pressure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Richard Wright

    2014-08-01

    ABSTRACT Evaluation of potential Reactor Pressure Vessel (RPV) steels has been carried out as part of the pre-conceptual Very High Temperature Reactor (VHTR) design studies. These design studies have generally focused on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Initially, three candidate materials were identified by this process: conventional light water reactor (LWR) RPV steels A508 and A533, 2¼Cr-1Mo in the annealed condition, and Grade 91 steel. The low strength of 2¼Cr-1Mo at elevated temperature has eliminated this steel from serious consideration as the VHTR RPV candidate material. Discussions with the very few vendors that can potentially produce large forgings for nuclear pressure vessels indicate a strong preference for conventional LWR steels. This preference is based in part on extensive experience with forging these steels for nuclear components. It is also based on the inability to cast large ingots of the Grade 91 steel due to segregation during ingot solidification, thus restricting the possible mass of forging components and increasing the amount of welding required for completion of the RPV. Grade 91 steel is also prone to weld cracking and must be post-weld heat treated to ensure adequate high-temperature strength. There are also questions about the ability to produce, and very importantly, verify the through thickness properties of thick sections of Grade 91 material. The availability of large components, ease of fabrication, and nuclear service experience with the A508 and A533 steels strongly favor their use in the RPV for the VHTR. Lowering the gas outlet temperature for the VHTR to 750°C from 950 to 1000°C, proposed in early concept studies, further strengthens the justification for this material selection. This steel is allowed in the ASME Boiler and Pressure Vessel Code for nuclear service up to 371°C (700°F); certain excursions above that temperature are

  9. Structure mechanical analysis of prestressed cast-steel pressure vessels with the finite-element-method

    International Nuclear Information System (INIS)

    Edalat, B.

    1981-08-01

    The analytical pressure analysis is performed for a vessel with solid bottom and top. The basis of the Finite-Element-Method (FEM) and the criteria for the choice of a suitable element type for use in the computer model was investigated. To investigate the exactness of the FE-program a comparison between the analytical solution and the pressure claculated by FEM at a cylindrical vessel was made. For pressure analyses at the test vessel built of steel sections four different computer models (after FEM) were developed. The pressure analysis of a prestressed cast-steel pressure vessel for the transport and for the storage of burnt HTR fuel elements is performed with the aid of computed models after FEM. The method of developing simple computer models for the prestressed pressure vessel with large dimension is explained with an example. (orig.) [de

  10. Test of 6-in. -thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C.; Smith, G.C.; Smith, J.E.; Whitman, G.D.

    1978-02-01

    HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7.

  11. Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7A under sustained loading

    International Nuclear Information System (INIS)

    Bryan, R.H.; Cate, T.M.; Holz, P.P.; King, T.A.; Merkle, J.G.; Robinson, G.C.; Smith, G.C.; Smith, J.E.; Whitman, G.D.

    1978-01-01

    HSST intermediate test vessel V-7 was repaired after being tested hydrostatically to leakage and was retested pneumatically as vessel V-7A. Except for the method of applying the load, the conditions in both tests were nearly identical. In each case, a sharp outside surface flaw 547 mm long (18 in.) by about 135 mm deep (5.3 in.) was prepared in the 152-mm-thick (6-in.) test cylinder of A533, grade B, class 1 steel. The inside surface of vessel V-7A was sealed in the region of the flaw by a thin metal patch so that pressure could be sustained after rupture. Vessel V-7A failed by rupture of the flaw ligament without burst, as expected. Rupture occurred at 144.3 MPa (20.92 ksi), after which pressure was sustained for 30 min without any indication of instability. The rupture pressure of vessel V-7A was about 2 percent less than that of vessel V-7

  12. Effect of radiation damage on operating safety of steel pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Vacek, M.; Havel, S.; Stoces, B.; Brumovsky, M.

    1980-01-01

    The effects are assessed of the environment upon mechanical properties of steel used generally for pressure vessels of light water nuclear reactors. Changes caused by radiation affect the reliability of vessels. Deterioration of steel properties is mainly due to neutron radiation. The article deals with factors bearing upon damage and with methods allowing to evaluate the reliability of vessels and predict their service life. Operating reliability of vessels is very unfavourably affected by planned and accidental reactor transients. (author)

  13. Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhou Dan; Zhuang Dongzhen

    2013-01-01

    The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

  14. Renovation of the sealing planes of WWER-400 reactors pressure vessel

    International Nuclear Information System (INIS)

    Jablonicky, P.; Pilat, P.

    2007-01-01

    An article describes technical solution for renovation of the sealing planes of WWER-440 reactor's pressure vessel. Four nickel sealing rings placed in four concentric grooves are providing hermetic sealing between the vessel and the lid of this type of the reactor. Impeccable seal of the reactor's pressure vessel, where the fission reaction takes place, represents a basic security factor for safe electric energy production. Principle of renovation of the reactor's pressure vessel and lid sealing planes is based on mechanical enlargement of defective grooves and following cladding of the new material by TIG welding. Final step for renovation includes machining of new grooves according to geometrical and surface quality requirements (Authors)

  15. Fabrication techniques of metal liner used for pressure vessels made by composite material

    International Nuclear Information System (INIS)

    Takahashi, W.K.; Al-Qureshi, H.A.

    1982-01-01

    Different viable techniques for the manufacturing of metal liner used for pressure vessels are presented. The aim of these metal liner is to avoid the fluid leakage from the pressurized vessel and to serve as a mandreal to be wound by composite material. The studied techniques are described and the practical results are illustrated. Finally a comparative study of the manufacturing techniques is made in order to define the process that furnishes the metal liner with the best characteristics. The advantages offered by these type of pressure vessels when compared with the conventional metallic vessels, are also presented. (Author) [pt

  16. A simplified approach for assessing the leak-before-break for the flawed pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kannan, P. [Ramagundam Super Thermal Power Station, NTPC Ltd, Jyothinagar 505215 (India); Amirthagadeswaran, K.S. [Faculty of Mechanical Engineering, Government College of Technology, Coimbatore 641013 (India); Christopher, T. [Faculty of Mechanical Engineering, Government College of Engineering, Tirunelveli 627007 (India); Nageswara Rao, B., E-mail: bnrao52@rediffmail.com [Faculty of Mechanical Engineering, School of Mechanical and Civil Sciences, K L University, Green Fields, Vaddeswaram, Guntur 522502 (India)

    2016-06-15

    Surface cracks or embedded cracks in pressure vessels under service may grow and form stable through-thickness cracks causing leak prior to failure. If this leak-before-break phenomenon takes place, then there is a possibility of preventing the vessel failure. This paper presents a simplified approach for assessing the leak-before-break or failure of the flawed pressure vessels. This approach is validated through comparison of existing test data.

  17. Guiding device for a manipulator mast for internal inspection of a reactor pressure vessel

    International Nuclear Information System (INIS)

    Seifert, W.; Schlueter, H.

    1977-01-01

    A remote-controlled supporting device centering a manipulator mast is described which is mounted and operated above a reactor pressure vessel under water in such a way that rotations and vertical movements necessary for the internal inspection of the pressure vessel remain possible. (RW) [de

  18. 30 CFR 56.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and... METAL AND NONMETAL MINES Compressed Air and Boilers § 56.13015 Inspection of compressed-air receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure vessels...

  19. Advances in crack-arrest technology for reactor pressure vessels

    International Nuclear Information System (INIS)

    Bass, B.R.; Pugh, C.E.

    1988-01-01

    The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs

  20. Underwater television camera for monitoring inner side of pressure vessel

    International Nuclear Information System (INIS)

    Takayama, Kazuhiko.

    1997-01-01

    An underwater television support device equipped with a rotatable and vertically movable underwater television camera and an underwater television camera controlling device for monitoring images of the inside of the reactor core photographed by the underwater television camera to control the position of the underwater television camera and the underwater light are disposed on an upper lattice plate of a reactor pressure vessel. Both of them are electrically connected with each other by way of a cable to rapidly observe the inside of the reactor core by the underwater television camera. The reproducibility is extremely satisfactory by efficiently concentrating the position of the camera and image information upon inspection and observation. As a result, the steps for periodical inspection can be reduced to shorten the days for the periodical inspection. Since there is no requirement to withdraw fuel assemblies over a wide reactor core region, and the device can be used with the fuel assemblies being left as they are in the reactor, it is suitable for inspection of detectors for nuclear instrumentation. (N.H.)

  1. Shallow-crack toughness results for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Theiss, T.J.; Shum, D.K.M.; Rolfe, S.T.

    1992-01-01

    The Heavy Section Steel Technology Program (HSST) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. To complete this investigation, techniques were developed to determine the fracture toughness from shallow-crack specimens. A total of 38 deep and shallow-crack tests have been performed on beam specimens about 100 mm deep loaded in 3-point bending. Two crack depths (a ∼ 50 and 9 mm) and three beam thicknesses (B ∼ 50, 100, and 150 mm) have been tested. Techniques were developed to estimate the toughness in terms of both the J-integral and crack-tip opening displacement (CTOD). Analytical J-integral results were consistent with experimental J-integral results, confirming the validity of the J-estimation schemes used and the effect of flaw depth on fracture toughness. Test results indicate a significant increase in the fracture toughness associated with the shallow flaw specimens in the lower transition region compared to the deep-crack fracture toughness. There is, however, little or no difference in toughness on the lower shelf where linear-elastic conditions exist for specimens with either deep or shallow flaws. The increase in shallow-flaw toughness compared with deep-flaw results appears to be well characterized by a temperature shift of 35 degree C

  2. Embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1995-01-01

    The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is i good surrogate for shift recovery and that there is a high level of consistency between he observed annealing trends and fundamental models of embrittlement and recovery processes

  3. Embrittlement recovery due to annealing of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Eason, E.D.; Wright, J.E.; Nelson, E.E.; Odette, G.R.; Mader, E.V.

    1998-01-01

    The irradiation embrittlement of nuclear reactor pressure vessels (RPV) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. The objective of this work was to analyze the pertinent data and develop quantitative models for estimating the recovery in 41 J (30 ft-lb) Charpy transition temperature (TT) and Charpy upper shelf energy (USE) due to annealing. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Models were developed based on a combination of statistical techniques, including pattern recognition and transformation analysis, and the current understanding of the mechanisms governing embrittlement and recovery. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and a surrogate hardness data base. This work demonstrates that microhardness recovery is a good surrogate for shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes. (orig.)

  4. Radiation-induced embrittlement in light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.; McElroy, W.N.

    1987-01-01

    In operating light water reactor (LWR) commercial power plants, neutron radiation induces embrittlement of the pressure vessel (PV) and its support structures. As a consequence, LWR-PV integrity is a primary safety consideration. LWR-PV integry is a significant economic consideration because the PV and its support structures are nonreplaceable power plant components and embrittlement of these components can therefore limit the effective operating lifetime of the plant. In addition to plant life considerations, LWR-PV embrittlement creates significant cycle-to-cycle impact through the restriction of normal heat-up and cool-down reactor operations. Recent LWR-PV benchmark experiments are analyzed. On this bases, it is established that an exponential representation accurately describes the spatial dependence of neutron exposure in LWR-PV. Implications produced by simple exponental behavior are explained and trend-curve models for the predictions of PV embrittelment are derived. These derivations provide for a clearer understanding and assessment of the assumptions underlying these trend-curve models. It is demonstrated that LWR-PV embrittlement possesses significant material dependence. (orig.)

  5. Break location influence in pressure vessel SBLOCA scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Querol, Andrea; Gallardo, Sergio; Verdú, Gumersindo, E-mail: anquevi@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain)

    2017-07-01

    The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused in simulating different hypothetical accidental scenarios in the PV using the thermalhydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios. (author)

  6. Development of automatic Ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Kim, Kor R.; Kim, Jae H.; Lee, Jae C.

    1996-06-01

    The selected weld areas of a reactor pressure vessel and adjacent piping are examined by the remote mechanized ultrasonic testing (MUT) equipment. Since the MUT equipment was purchased from southwest Research Institute (SwRI) in April 1985, 15 inservice inspections and 5 preservice inspections are performed with this MUT equipment. However due to the old age of the equipment and frequent movements to plant sites, the reliability of examination was recently decreased rapidly and it is very difficult to keep spare parts. In order to resolve these problems and to meet the strong request from plant sites, we intend to develop a new 3-axis control system including hardware and software. With this control system, we expect more efficient and reliable examination of the nozzle to shell weld areas, which is specified in ASME Code Section XI. The new 3-axis control system hardware and software were designed and development of our own control system, the advanced technologies of computer control mechanism were established and examination reliability of the nozzle to shell weld area was improved. With the development of our 3-axis control system for PaR ISI-2 computer control system, the reliability of nozzle to shell weld area examination has been improved. The established technologies from the development and detailed analysis of existing control system, are expected to be applied to the similar control systems in nuclear power plants. (author). 12 refs., 4 tabs., 33 figs

  7. Updated embrittlement trend curve for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kirk, M.; Santos, C.; Eason, E.; Wright, J.; Odette, G.R.

    2003-01-01

    The reactor pressure vessels of commercial nuclear power plants are subject to embrittlement due to exposure to high energy neutrons from the core. Irradiation embrittlement of RPV belt-line materials is currently evaluated using US Regulatory Guide 1.99 Revision 2 (RG 1.99 Rev 2), which presents methods for estimating the Charpy transition temperature shift (ΔT30) at 30 ft-lb (41 J) and the drop in Charpy upper shelf energy (ΔUSE). A more recent embrittlement model, based on a broader database and more recent research results, is presented in NUREG/CR-6551. The objective of this paper is to describe the most recent update to the embrittlement model in NUREG/CR-6551, based upon additional data and increased understanding of embrittlement mechanisms. The updated ΔT30 and USE models include fluence, copper, nickel, phosphorous content, and product form; the ΔT30 model also includes coolant temperature, irradiation time (or flux), and a long-time term. The models were developed using multi-variable surface fitting techniques, understanding of the ΔT30 mechanisms, and engineering judgment. The updated ΔT30 model reduces scatter significantly relative to RG 1.99 Rev 2 on the currently available database for plates, forgings, and welds. This updated embrittlement trend curve will form the basis of revision 3 to Regulatory Guide 1.99. (author)

  8. Miniaturized Charpy test for reactor pressure vessel embrittlement characterization

    Energy Technology Data Exchange (ETDEWEB)

    Manahan, M.P. Sr. [MPM Research and Consulting, Lemont, PA (United States)

    1999-10-01

    Modifications were made to a conventional Charpy machine to accommodate the miniaturized Charpy V-Notch (MCVN) specimens which were fabricated from an archived reactor pressure vessel (RPV) steel. Over 100 dynamic MCVN tests were performed and compared to the results from conventional Charpy V-Notch (CVN) tests to demonstrate the efficacy of the miniature specimen test. The optimized sidegrooved MCVN specimens exhibit transitional fracture behavior over essentially the same temperature range as the CVN specimens which indicates that the stress fields in the MCVN specimens reasonably simulate those of the CVN specimens and this fact has been observed in finite element calculations. This result demonstrates a significant breakthrough since it is now possible to measure the ductile-brittle transition temperature (DBTT) using miniature specimens with only small correction factors, and for some materials as in the present study, without the need for any correction factor at all. This development simplifies data interpretation and will facilitate future regulatory acceptance. The non-sidegrooved specimens yield energy-temperature data which is significantly shifted downward in temperature (non-conservative) as a result of the loss of constraint which accompanies size reduction.

  9. Development of automatic ultrasonic testing equipment for reactor pressure vessel

    International Nuclear Information System (INIS)

    Jang, Kee Ok; Park, Dae Yung; Park, Moon Hoh; Koo, Kil Mo; Park, Kwang Heui; Kang, Sang Sin; Bang, Heui Song; Noh, Heui Choong; Kong, Woon Sik

    1994-08-01

    The selected weld areas of reactor pressure vessel and adjacent piping are examined by remote mechanized ultrasonic testing(MUT) equipment. Since the MUT equipment was purchased from Southwest Research Institute (SwRI) in April 1985, we have performed 15 inservice inspections and 5 preservice inspections. However, the reliability of examination was recently decreased rapidly as the problems which results from the old age of equipment and the frequent movement to plant site to site have occurred frequently. Therefore, the 3-axis control system hardware in occurring many problems among the equipments of mechanized ultrasonic testing (MUT) was designed and developed to cover the examination areas of nozzle-shell weld as specified in ASME Code Section XI and to improve the examination reliability. The new 3-axis control system hardware with the performance of this project was developed to be compatible with the old one and it was used as dual system or spare parts of the old system. Furthermore, the established technologies are expected to be applied to the similar control systems in nuclear power plant. 17 figs, 2 pix, 2 tabs, 10 refs. (Author)

  10. Neutron irradiation effects on spark plasma sintered boron carbide

    International Nuclear Information System (INIS)

    Buyuk, B.; Cengiz, M.; Tugrul, A.; Ozer, S.; Yucel, O.; Goller, G.; Sahin, F.C.; Lastovski, S.B.

    2015-01-01

    In this study, spark plasma sintered boron carbide (B 4 C) was examined against neutrons. The specimens were ir-radiated by reactor neutrons (include both thermal and fast neutrons) up to fluence of 1.37x10 21 n*m -2 . Thermal and fast neutrons cause swelling by different interactions with boron ( 10 B) atoms in the related materials. X-Ray diffraction (XRD) patterns and scanning electron microscopy (SEM) images were investigated for initial and irradiated samples. In addition, lattice parameters and unit cell volumes were calculated for the samples. The swelling percentages were calculated to be within a range of 0.49-3.80 % (average 1.70 %) for the outer surface of the materials for applied neutron irradiation doses. (authors)

  11. Tensile and fracture toughness test results of neutron irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

    1998-01-01

    Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

  12. Effect of neutron irradiation on p-type silicon

    International Nuclear Information System (INIS)

    Sopko, B.

    1973-01-01

    The possibilities are discussed of silicon isotope reactions with neutrons of all energies. In the reactions, 30 Si is converted to a stable phosphorus isotope forming n-type impurities in silicon. The above reactions proceed as a result of thermal neutron irradiation. An experiment is reported involving irradiation of two p-type silicon single crystals having a specific resistance of 2000 ohm.cm and 5000 to 20 000 ohm.cm, respectively, which changed as a result of irradiation into n-type silicon with a given specific resistance. The specific resistance may be pre-calculated from the concentration of impurities and the time of irradiation. The effects of irradiation on other silicon parameters and thus on the suitability of silicon for the manufacture of semiconductor elements are discussed. (J.K.)

  13. Preparation of 227Ac by neutron irradiation of 226Ra

    International Nuclear Information System (INIS)

    Kukleva, E.; Kozempel, J.; Vlk, M.; Micolova, P.; Vopalka, D.

    2015-01-01

    Radium-223 is prospective alpha-emitting therapeutic radionuclide for targeted radionuclide therapy. Although 223 Ra is formed naturally by the decay of 235 U, for practical reasons its preparation involves neutron irradiation of 226 Ra. The α-decay of the 227 Ra (T 12 = 43 min.) produced via 226 Ra(n,γ) 227 Ra reaction leads to 227 Ac, a mother nuclide of 227 Th and 223 Ra subsequently. Irradiation target radium material is generally available in multi-gram quantities from historical stock. Main aim of this study was to experimentally and theoretically evaluate and verify available literature data on production of 223 Ra. According to data obtained from γ-spectra, the approximate yield values were determined and effective cross-section for the 223 Ra production was calculated. (authors)

  14. Neutron irradiation and compatibility testing of Li2O

    International Nuclear Information System (INIS)

    Porter, D.L.; Krsul, J.R.; Laug, M.T.; Walters, L.C.; Tetenbaum, M.

    1983-01-01

    A study was made of the neutron-irradiation behavior of 6 Li-enriched Li 2 O material in EBR-II. In addition, a stress-corrosion study was performed ex-reactor to test compatibility of Li 2 O materials with a variety of stainless steels. Results of the irradiation testing showed that tritium and helium retention in the Li 2 O (approx. 89% dense) lessened with neutron exposure. Helium and tritium retention appear to approach steady-state after approx. 1% 6 Li burnup. The stress-corrosion studies, using 316 stainless steel (Ti-modified) and a 35% Ni alloy, showed that stress does not enhance the corrosion, and that dry Li 2 O is not significantly corrosive, the LiOH content producing the corrosive effects. Corrosion, in general, was not severe as a passivation in sealed capsules seemed to occur after a time, greatly reducing corrosion rates

  15. ATF Neutron Irradiation Program Irradiation Vehicle Design Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Geringer, J. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Cetiner, N. O. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division; McDuffee, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Materials Science and Technology Division

    2016-03-01

    The Japan Atomic Energy Agency (JAEA) under the Civil Nuclear Energy Working Group (CNWG) is engaged in a cooperative research effort with the U.S. Department of Energy (DOE) to explore issues related to nuclear energy, including research on accident-tolerant fuels and materials for use in light water reactors. This work develops a draft technical plan for a neutron irradiation program on the candidate accident-tolerant fuel cladding materials and elements using the High Flux Isotope Reactor (HFIR). The research program requires the design of a detailed experiment, development of test vehicles, irradiation of test specimens, possible post irradiation examination and characterization of irradiated materials and the shipment of irradiated materials to Japan. This report discusses the conceptual design, the development and irradiation of the test vehicles.

  16. Design, Construction, and Modeling of a 252Cf Neutron Irradiator

    Directory of Open Access Journals (Sweden)

    Blake C. Anderson

    2016-01-01

    Full Text Available Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

  17. Proceedings of neutron irradiation technical meeting on BNCT

    International Nuclear Information System (INIS)

    2000-10-01

    The 'Neutron Irradiation Technical Meeting for Boron Neutron Capture Therapy (BNCT)' was held on March 13, 2000 at Tokai Research Establishment. The Meeting is aimed to introduce the neutron beam facility for medical irradiation at JRR-4 to Japanese researchers widely, as well as providing an opportunity for young researchers, engineers, medical representatives such surgeons and doctors of pharmacology to present their research activities and to exchange valuable information. JAERI researcher presented the performance and the irradiation technology in the JRR-4 neutron beam facility, while external researchers made various and beneficial presentations containing such accelerator-based BNCT, spectrum-shifter, biological effect, pharmacological development and so on. In this meeting, a special lecture titled 'The Dawn of BNCT and Its Development.' was given by MD, Prof. Takashi Minobe, an executive director of Japan Foundation for Emergency Medicine. The 11 of the presented papers are indexed individually. (J.P.N.)

  18. Proceedings of neutron irradiation technical meeting on BNCT

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

    The 'Neutron Irradiation Technical Meeting for Boron Neutron Capture Therapy (BNCT)' was held on March 13, 2000 at Tokai Research Establishment. The Meeting is aimed to introduce the neutron beam facility for medical irradiation at JRR-4 to Japanese researchers widely, as well as providing an opportunity for young researchers, engineers, medical representatives such surgeons and doctors of pharmacology to present their research activities and to exchange valuable information. JAERI researcher presented the performance and the irradiation technology in the JRR-4 neutron beam facility, while external researchers made various and beneficial presentations containing such accelerator-based BNCT, spectrum-shifter, biological effect, pharmacological development and so on. In this meeting, a special lecture titled 'The Dawn of BNCT and Its Development.' was given by MD, Prof. Takashi Minobe, an executive director of Japan Foundation for Emergency Medicine. The 11 of the presented papers are indexed individually. (J.P.N.)

  19. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Sone, Kazuho; Shiraishi, Kensuke

    1976-01-01

    Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

  20. Microstructural defects in EUROFER 97 after different neutron irradiation conditions

    Directory of Open Access Journals (Sweden)

    Christian Dethloff

    2016-12-01

    Full Text Available Characterization of irradiation induced microstructural evolution is essential for assessing the applicability of structural steels like the Reduced Activation Ferritic/Martensitic steel EUROFER 97 in upcoming fusion reactors. In this work Transmission Electron Microscopy (TEM is used to determine the defect microstructure after different neutron irradiation conditions. In particular dislocation loops, voids and precipitates are analyzed concerning defect nature, density and size distribution after irradiation to 15 dpa at 300 °C in the mixed spectrum High Flux Reactor (HFR. New results are combined with previously obtained data from irradiation in the fast spectrum BOR-60 reactor (15 and 32 dpa, 330 °C, which allows for assessment of dose and dose rate effects on the aforementioned irradiation induced defects and microstructural characteristics.

  1. Germanium-doped gallium phosphide obtained by neutron irradiation

    Science.gov (United States)

    Goldys, E. M.; Barczynska, J.; Godlewski, M.; Sienkiewicz, A.; Heijmink Liesert, B. J.

    1993-08-01

    Results of electrical, optical, electron spin resonance and optically detected magnetic resonance studies of thermal neutron irradiated and annealed at 800 °C n-type GaP are presented. Evidence is found to support the view that the main dopant introduced via transmutation of GaP, germanium, occupies cation sites and forms neutral donors. This confirms the possibility of neutron transmutation doping of GaP. Simultaneously, it is shown that germanium is absent at cation sites. Presence of other forms of Ge-related defects is deduced from luminescence and absorption data. Some of them are tentatively identified as VGa-GeGa acceptors leading to the self-compensation process. This observation means that the neutron transmutation as a doping method in application to GaP is not as efficient as for Si.

  2. Irradiation hardening of pure tungsten exposed to neutron irradiation

    Science.gov (United States)

    Hu, Xunxiang; Koyanagi, Takaaki; Fukuda, Makoto; Kumar, N. A. P. Kiran; Snead, Lance L.; Wirth, Brian D.; Katoh, Yutai

    2016-11-01

    Pure tungsten samples have been neutron irradiated in HFIR at 90-850 °C to 0.03-2.2 dpa. A dispersed barrier hardening model informed by the available microstructure data has been used to predict the hardness. Comparison of the model predictions and the measured Vickers hardness reveals the dominant hardening contribution at various irradiation conditions. For tungsten samples irradiated in HFIR, the results indicate that voids and dislocation loops contributed to the hardness increase in the low dose region (0.6 dpa). The precipitate contribution is most pronounced for the HFIR irradiations, whereas the radiation-induced defect cluster microstructure can rationalize the entirety of the hardness increase observed in tungsten irradiated in the fast neutron spectrum of Joyo and the mixed neutron spectrum of JMTR.

  3. Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

    Directory of Open Access Journals (Sweden)

    Hereil Pierre-Louis

    2015-01-01

    Full Text Available Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

  4. Results of neutron irradiation of liquid lithium saturated with deuterium

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibayeva, Irina, E-mail: tazhibayeva@ntsc.kz [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Ponkratov, Yuriy; Kulsartov, Timur; Gordienko, Yuriy; Skakov, Mazhyn; Zaurbekova, Zhanna [Institute of Atomic Energy, National Nuclear Center of RK, Kurchatov (Kazakhstan); Lyublinski, Igor [JSC «Red Star», Moscow (Russian Federation); NRNU «MEPhI», Moscow (Russian Federation); Vertkov, Alexey [JSC «Red Star», Moscow (Russian Federation); Mazzitelli, Giuseppe [ENEA, RC Frascati, Frascati (Italy)

    2017-04-15

    Highlights: • The results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1M research reactor are described. • At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(−144/RT). • The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise. - Abstract: This paper describes the results on neutron irradiation of liquid lithium saturated with deuterium at the IVG.1 M research reactor. The neutron flux at the reactor core center at 2 MW was 5 10{sup −13} cm{sup −2} s{sup −1}. The efficiency coefficients of helium and tritium release from lithium saturated with deuterium were calculated. The tritium interaction with lithium atoms (formation and dissociation of lithium tritide) has an effect on tritium release. An increment of sample’s temperature results in tritium release acceleration due to rising of the dissociation rate of lithium tritide. At temperatures below 573 K the efficiency coefficient of tritium release is well described by the expression K = 0.015 exp(−14/RT), and above 623 K − K = 10{sup 9} exp(-144/RT). The T{sub 2} molecules contribution into the overall tritium release becomes apparent at temperatures higher than 673 K and increases with the temperature rise.

  5. Neutron irradiation effects on AlGaN/GaN high electron mobility transistors

    International Nuclear Information System (INIS)

    Lü Ling; Zhang Jin-Cheng; Xue Jun-Shuai; Ma Xiao-Hua; Zhang Wei; Bi Zhi-Wei; Zhang Yue; Hao Yue

    2012-01-01

    AlGaN/GaN high electron mobility transistors (HEMTs) were exposed to 1 MeV neutron irradiation at a neutron fluence of 1 × 10 15 cm −2 . The dc characteristics of the devices, such as the drain saturation current and the maximum transconductance, decreased after neutron irradiation. The gate leakage currents increased obviously after neutron irradiation. However, the rf characteristics, such as the cut-off frequency and the maximum frequency, were hardly affected by neutron irradiation. The AlGaN/GaN heterojunctions have been employed for the better understanding of the degradation mechanism. It is shown in the Hall measurements and capacitance—voltage tests that the mobility and concentration of two-dimensional electron gas (2DEG) decreased after neutron irradiation. There was no evidence of the full-width at half-maximum of X-ray diffraction (XRD) rocking curve changing after irradiation, so the dislocation was not influenced by neutron irradiation. It is concluded that the point defects induced in AlGaN and GaN by neutron irradiation are the dominant mechanisms responsible for performance degradations of AlGaN/GaN HEMT devices. (condensed matter: electronic structure, electrical, magnetic, and optical properties)

  6. Neutron irradiation effects on AlGaN/GaN high electron mobility transistors

    Science.gov (United States)

    Lü, Ling; Zhang, Jin-Cheng; Xue, Jun-Shuai; Ma, Xiao-Hua; Zhang, Wei; Bi, Zhi-Wei; Zhang, Yue; Hao, Yue

    2012-03-01

    AlGaN/GaN high electron mobility transistors (HEMTs) were exposed to 1 MeV neutron irradiation at a neutron fluence of 1 × 1015 cm-2. The dc characteristics of the devices, such as the drain saturation current and the maximum transconductance, decreased after neutron irradiation. The gate leakage currents increased obviously after neutron irradiation. However, the rf characteristics, such as the cut-off frequency and the maximum frequency, were hardly affected by neutron irradiation. The AlGaN/GaN heterojunctions have been employed for the better understanding of the degradation mechanism. It is shown in the Hall measurements and capacitance—voltage tests that the mobility and concentration of two-dimensional electron gas (2DEG) decreased after neutron irradiation. There was no evidence of the full-width at half-maximum of X-ray diffraction (XRD) rocking curve changing after irradiation, so the dislocation was not influenced by neutron irradiation. It is concluded that the point defects induced in AlGaN and GaN by neutron irradiation are the dominant mechanisms responsible for performance degradations of AlGaN/GaN HEMT devices.

  7. Determination of the critical buckling pressure of blood vessels using the energy approach.

    Science.gov (United States)

    Han, Hai-Chao

    2011-03-01

    The stability of blood vessels under lumen blood pressure is essential to the maintenance of normal vascular function. Differential buckling equations have been established recently for linear and nonlinear elastic artery models. However, the strain energy in bent buckling and the corresponding energy method have not been investigated for blood vessels under lumen pressure. The purpose of this study was to establish the energy equation for blood vessel buckling under internal pressure. A buckling equation was established to determine the critical pressure based on the potential energy. The critical pressures of blood vessels with small tapering along their axis were estimated using the energy approach. It was demonstrated that the energy approach yields both the same differential equation and critical pressure for cylindrical blood vessel buckling as obtained previously using the adjacent equilibrium approach. Tapering reduced the critical pressure of blood vessels compared to the cylindrical ones. This energy approach provides a useful tool for studying blood vessel buckling and will be useful in dealing with various imperfections of the vessel wall.

  8. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

  9. Studies on formation and structures of ultrafine Cu precipitates in Fe-Cu model alloys for reactor pressure vessel steels using positron quantum dot confinement in the precipitates by their positron affinity. JAERI's nuclear research promotion program, H11-034 (Contract research)

    CERN Document Server

    Hasegawa, M; Suzuki, M; Tang, Z; Yubuta, K

    2003-01-01

    Positron annihilation experiments on Fe-Cu model dilute alloys of nuclear reactor pressure vessel (RPV) steels have been performed after neutron irradiation in JMTR. Nanovoids whose inner surfaces were covered by Cu atoms were clearly observed. The nanovoids transformed to ultrafine Cu precipitates by dissociating their vacancies after annealing at around 400degC. The nanovoids and the ultrafine Cu precipitates are strongly suggested to be responsible for irradiation-induced embrittlement of RPV steels. Effects of Ni, Mn and P addition on the nanovoid and Cu precipitate formations were also studied. The nanovoid formation was enhanced by Ni and P, but suppressed by Mn. The Cu precipitates after annealing around 400degC were almost free from these doping elements and hence were pure Cu in the chemical composition. Furthermore the Fermi surface of the 'embedded' Cu precipitates with a body centered cubic crystal structure was obtained from two dimensional angular correlation of annihilation radiation (2D-ACAR) ...

  10. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel and outer shell around the primary pressure vessel and acting as a protection for it against outside projectiles. A floor is provided internally dividing the outside shell into two upper and lower sections and an inside wall dividing the lower section into one part containing the primary pressure vessel and a second part, both made pressure tight with respect to each other and with the outside shell and forming with the latter a secondary means of containment [fr

  11. Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

    International Nuclear Information System (INIS)

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  12. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  13. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  14. Reactors with pressure vessel in pre-stressed concrete

    International Nuclear Information System (INIS)

    Devillers, Christian; Lafore, Pierre

    1964-12-01

    After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

  15. Study of gem materials by neutron irradiation: characterization of impurities and color centers

    International Nuclear Information System (INIS)

    Leal, Alexandre S.; Menezes, Maria A.B.C.; Brito, Walter de; D'Urco, Ana F.A.; Felix, Marcia C.; Krambrock, Klaus; Ferreira, Ana F.

    2005-01-01

    Since one-century laboratory irradiation techniques are applied to the color enhancement of gem minerals. Its actual status and applications are discussed. Many different colors in a variety of gem minerals can be produced by gamma, electron and neutron irradiation combined with thermal treatments, however, many color centers and coloration processes are not known in detail. In this work we present examples of neutron irradiation applied to colorless topaz, spodumene and diamond. Topaz and diamond turned blue, spodumene orange. All color centers produced by neutron irradiation are stable to elevated temperatures and can be considered as color enhancing processes. (author)

  16. Simulation of Sark Current Increase in Si PIN Photodiode Induced by Neutron Irradiation

    International Nuclear Information System (INIS)

    Wang Zujun; Chen Wei; Zhang Yong; Tang Benqi; Xiao Zhigang; Huang Shaoyao; Liu Minbo; Liu Yinong

    2010-01-01

    The mechanism of dark current increase in Si PIN photodiode induced by neutron irradiation was analyzed. The device physics and neutron irradiation models were presented to simulate dark current in Si PIN photodiode by MEDICI software. The primary regularity of dark current increase in Si PIN photodiode was concluded by neutron irradiation with the energy of 1 MeV and at the fluence of 10 10 -10 14 cm -2 . The simulation results are in agreement with the experimental results from relevant literature. (authors)

  17. Shielding of a neutron irradiator with {sup 241}Am-Be source

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, K.A.M. de; Crispim, V.R.; Silva, A.X., E-mail: koliveira@con.ufrj.b, E-mail: verginia@con.ufrj.b, E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear; Fonseca, E.S., E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The equivalent dose rates at 1.0 cm from the outer surface of the shielding of a neutron irradiation system that uses {sup 241}Am-Be source with activity of 185 GBq (5 Ci) were determined. A theoretical-experimental approach including case studies, through computer simulations with MCNP code was employed to calculate the best shielding thickness. Following the construction of the neutron irradiator, dose measurements were conducted in order to validate data obtained from simulation. The neutron irradiator shielding was designed in such a way to allow transport of the neutron radiography system for in loco inspections ensuring workers' radiologic safety. (author)

  18. Some Observations on Damage Tolerance Analyses in Pressure Vessels

    Science.gov (United States)

    Raju, Ivatury S.; Dawicke, David S.; Hampton, Roy W.

    2017-01-01

    AIAA standards S080 and S081 are applicable for certification of metallic pressure vessels (PV) and composite overwrap pressure vessels (COPV), respectively. These standards require damage tolerance analyses with a minimum reliable detectible flaw/crack and demonstration of safe life four times the service life with these cracks at the worst-case location in the PVs and oriented perpendicular to the maximum principal tensile stress. The standards require consideration of semi-elliptical surface cracks in the range of aspect ratios (crack depth a to half of the surface length c, i.e., (a/c) of 0.2 to 1). NASA-STD-5009 provides the minimum reliably detectible standard crack sizes (90/95 probability of detection (POD) for several non-destructive evaluation (NDE) methods (eddy current (ET), penetrant (PT), radiography (RT) and ultrasonic (UT)) for the two limits of the aspect ratio range required by the AIAA standards. This paper tries to answer the questions: can the safe life analysis consider only the life for the crack sizes at the two required limits, or endpoints, of the (a/c) range for the NDE method used or does the analysis need to consider values within that range? What would be an appropriate method to interpolate 90/95 POD crack sizes at intermediate (a/c) values? Several procedures to develop combinations of a and c within the specified range are explored. A simple linear relationship between a and c is chosen to compare the effects of seven different approaches to determine combinations of aj and cj that are between the (a/c) endpoints. Two of the seven are selected for evaluation: Approach I, the simple linear relationship, and a more conservative option, Approach III. For each of these two Approaches, the lives are computed for initial semi-elliptic crack configurations in a plate subjected to remote tensile fatigue loading with an R-ratio of 0.1, for an assumed material evaluated using NASGRO (registered 4) version 8.1. These calculations demonstrate

  19. Residual stresses in a weldment of pressure vessel steel

    International Nuclear Information System (INIS)

    Gott, K.E.

    1978-01-01

    A study was made of the distribution of residual stresses around a typical weld from a light water reactor pressure vessel by an X-ray double-exposure camera technique. So that the magnitude, sign, and distribution of the residual stresses were as similar as possible to those found in practice, a wide, full-thickness specimen of A533B Cl 1 steel containing a submerged-arc weld was stress-relief annealed. To obtain a three-dimensional distribution of the stresses the specimen was examined at different levels through the thickness. Following the removal of material by milling, the specimen surface was electropolished to free it from cold work. Corrections have been made to take into account specimen relaxation. To completely define the original stress system it is desirable also to measure the change in curvature on removing a layer of material. Unless this is done assumptions must be made which complicate the calculations unnecessarily. This became apparent after the experimental work was completed. In the centre of the plate the methods of correction which can be used are sensitive to errors in the measurements. The corrected results show that the dominant residual stress is perpendicular to the weld. It is positive at the surfaces and negative in the centre of the plate. The maximum value can reach the yield stress. The residual stresses in the weld metal can locally vary considerably: from 100 to 350N/mm 2 over a distance of 5mm. Such large variations have been found to coincide with the heat-affected zones of the individual weld runs. (author)

  20. UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

    International Nuclear Information System (INIS)

    Watson, P.S.

    1990-01-01

    Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

  1. On the Adequacy of API 521 Relief-Valve Sizing Method for Gas-Filled Pressure Vessels Exposed to Fire

    DEFF Research Database (Denmark)

    Andreasen, Anders; Nieto, Marcos Zan; Borroni, Filippo

    2018-01-01

    sense of security. Often the vessel wall will be weakened by high temperatures, before the PRV relieving pressure is reached. In this article, a multiparameter study has been performed taking into consideration various vessel sizes, design pressures (implicitly vessel wall thickness), vessel operating...

  2. Investigation of the failure of a reactor pressure vessel by plastic instability

    International Nuclear Information System (INIS)

    Laemmer, H.; Ritter, B.

    1994-01-01

    A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

  3. Development of computational methods of design by analysis for pressure vessel components

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan; Wu Honglin

    2005-01-01

    Stress classification is not only one of key steps when pressure vessel component is designed by analysis, but also a difficulty which puzzles engineers and designers at all times. At present, for calculating and categorizing the stress field of pressure vessel components, there are several computation methods of design by analysis such as Stress Equivalent Linearization, Two-Step Approach, Primary Structure method, Elastic Compensation method, GLOSS R-Node method and so on, that are developed and applied. Moreover, ASME code also gives an inelastic method of design by analysis for limiting gross plastic deformation only. When pressure vessel components design by analysis, sometimes there are huge differences between the calculating results for using different calculating and analysis methods mentioned above. As consequence, this is the main reason that affects wide application of design by analysis approach. Recently, a new approach, presented in the new proposal of a European Standard, CEN's unfired pressure vessel standard EN 13445-3, tries to avoid problems of stress classification by analyzing pressure vessel structure's various failure mechanisms directly based on elastic-plastic theory. In this paper, some stress classification methods mentioned above, are described briefly. And the computational methods cited in the European pressure vessel standard, such as Deviatoric Map, and nonlinear analysis methods (plastic analysis and limit analysis), are depicted compendiously. Furthermore, the characteristics of computational methods of design by analysis are summarized for selecting the proper computational method when design pressure vessel component by analysis. (authors)

  4. The influence of chemistry concentration on the fracture risk of a reactor pressure vessel subjected to pressurized thermal shocks

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Pin-Chiun [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China); Chou, Hsoung-Wei, E-mail: hwchou@iner.gov.tw [Institute of Nuclear Energy Research, Taoyuan 32546, Taiwan, ROC (China); Ferng, Yuh-Ming [Institute of Nuclear Engineering and Science, National Tsing-Hua University, Hsinchu 30013, Taiwan, ROC (China)

    2016-02-15

    Highlights: • Probabilistic fracture mechanics method was used to analyze a reactor pressure vessel. • Effects of copper and nickel contents on RPV fracture probability under PTS were investigated and discussed. • Representative PTS transients of Beaver Valley nuclear power plant were utilized. • The range of copper and nickel contents of the RPV materials were suggested. • With different embrittlement levels the dominated PTS category is different. - Abstract: The radiation embrittlement behavior of reactor pressure vessel shell is influenced by the chemistry concentration of metal materials. This paper aims to study the effects of copper and nickel content variations on the fracture risk of pressurized water reactor (PWR) pressure vessel subjected to pressurized thermal shock (PTS) transients. The probabilistic fracture mechanics (PFM) code, FAVOR, which was developed by the Oak Ridge National Laboratory in the United States, is employed to perform the analyses. A Taiwan domestic PWR pressure vessel assumed with varied copper and nickel contents of beltline region welds and plates is investigated in the study. Some PTS transients analyzed from Beaver Valley Unit 1 for establishing the U.S. NRC's new PTS rule are applied as the loading condition. It is found that the content variation of copper and nickel will significantly affect the radiation embrittlement and the fracture probability of PWR pressure vessels. The results can be regarded as the risk incremental factors for comparison with the safety regulation requirements on vessel degradation as well as a reference for the operation of PWR plants in Taiwan.

  5. Change in properties of superconducting magnet materials by fusion neutron irradiation

    International Nuclear Information System (INIS)

    Nishimura, Arata; Nishijima, Shigehiro; Takeuchi, Takao; Nishitani, Takeo

    2007-01-01

    A fusion reactor will generate a lot of high energy neutron and much energy will be taken out of the neutrons by a blanket system. Since some neutrons will stream out of a plasma vacuum vessel through neutral beam injection ports and penetrate a blanket system, a superconducting magnet system, which provides high magnetic field to confirm high energy particles, will be irradiated by a certain amount of neutrons. By developing the new NBI system or by reducing the penetration, the neutron fluence to the superconducting magnet will be able to be reduced. However, it is not easy to achieve the lower streaming and penetration at the present. Therefore, investigations on irradiation behavior of superconducting magnet materials are desired and some novel researches have been performed from 1970s. In general, the critical current of the superconducting wire increases under fast neutron environment comparing with that of the non-irradiated wire, and then decreased to almost zero as an increase of neutron fluence. On the other hand, the critical temperature of the wire starts to get down around 10 22 n/m 2 of neutron fluence and the temperature margin will be decreased during the operation by the neutron irradiation. In this paper, some aspects of irradiated materials will be overviewed and general tendency will be discussed focussing on knock-on effect of fast neutron and long range ordering of A15 compounds

  6. Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

    International Nuclear Information System (INIS)

    Ventura, Mirta A.; Rosso, Ricardo D.

    2002-01-01

    In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

  7. An integrity evaluation method of the pressure vessel of nuclear reactors under pressurized thermal shock

    International Nuclear Information System (INIS)

    Matsubara, Masaaki; Okamura, Hiroyuki.

    1987-01-01

    Present paper proposes a new algorithm of the integrity evaluation of the pressure vessel of nuclear reactors under pressurized thermal shock, PTS. This method enables us to do an effective evaluation by superimposing proposed ''PTS state-transient curves'' and ''toughness transient curves'', and is superior to a conventional one in the following points; (1) easy to get an overall view of the result of PTS event for the variations of several parameters, (2) possible to evaluate a safety margin for irradiation embrittlement, and (3) enable to construct an Expert-friendly evaluation system. In addition, the paper shows that we can execute a safety assurance test by using a flat plate model with the same thickness as that of real plant. (author)

  8. Review of pressurized thermal shock studies of large scale reactor pressure vessels in Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Fekete

    2016-03-01

    Full Text Available In Hungary, four nuclear power units were constructed more than 30 years ago; they are operating to this day. In every unit, VVER-440 V213-type light-water cooled, light-water moderated, ressurized water reactors are in operation. Since the mid-1980s, numerous researches in the field of Pressurized Thermal Shock (PTS analyses of Reactor Pressure Vessels (RPVs have been conducted in Hungary; in all of them, the concept of structural integrity was the basis of research and development. During this time, four large PTS studies with industrial relevance have been completed in Hungary. Each used different objectives and guides, and the analysis methodology was also changing. This paper gives a comparative review of the methodologies used in these large PTS Structural Integrity Analysis projects, presenting the latest results as well

  9. Mössbauer spectroscopy of tin in unirradiated and neutron irradiated Zircaloys

    Science.gov (United States)

    Sawicki, Jerzy A.

    1999-01-01

    Mössbauer spectroscopy of the 23.9 keV γ-rays in 119Sn nuclei was applied to study Zircaloy-2, Zircaloy-4, and other tin-bearing zirconium-based alloys of interest to nuclear power technology. Zircaloys are extensively used in nuclear reactors as fuel cladding. In CANDU reactors, Zircaloys are also used as major structural components such as calandria tubes, and were used until the late 1970's as pressure tubes (now replaced by Zr-2.5Nb alloy). Unirradiated specimens of these alloys, as well as radioactive specimens, both neutron-irradiated in high-flux test reactors and extracted from nuclear power-reactor components after many years of service, were examined. The obtained spectra consistently showed tin in substitutional solid solution in α-Zr, whereas no evidence was found of metallic Sn or intermetallic Zr 4Sn precipitates. In oxide scrapes removed from Zircaloy-2 pressure tube of one of CANDU reactors, where the alloy was exposed for about 10 years to pressurized heavy water coolant at temperatures of ˜280°C, a considerable fraction of tin was found in the Sn(IV) state, in the form that coincides with the state of tin in stannic oxide, SnO 2. The same form of tin was identified in filterable deposits in the primary heavy water coolant of CANDU reactors. For comparison, in Zircaloy heated in air, SnO 2 was formed only at temperatures above 500°C.

  10. Preliminary study of an expert system for mechanical design of a pressure vessel

    International Nuclear Information System (INIS)

    Kasmuri, N.H.; Md Som, A.

    2006-01-01

    This paper describes a preliminary study of an expert system for mechanical design of a pressure vessel. The system supports the framework for the conceptual mechanical design from the initial stages within the design procedures. ASME Boiler and Pressure Vessel Code Section VIII Division 1 were applied as a design rule. The proposed methodology facilitates the development of knowledge base acquisition, knowledge base construction and the prototype implementation. This study characterizes a knowledge base (procedure) of mechanical design of a pressure vessel subjected to internal pressure including all design parameters; i.e. temperature, shell thickness, selection of materials of constructions, stress analysis procedure, support and ancillary items. The rationalization of the mechanical design is shown in the form of a schematic flow diagram. A Kappa PC expert system shell is used as a tool to develop the prototype software. It provides graphical representation for creating objects, hierarchies and rules for knowledge base used in pressure vessel design. (Author)

  11. Evaluation of Progressive Failure Analysis and Modeling of Impact Damage in Composite Pressure Vessels

    Science.gov (United States)

    Sanchez, Christopher M.

    2011-01-01

    NASA White Sands Test Facility (WSTF) is leading an evaluation effort in advanced destructive and nondestructive testing of composite pressure vessels and structures. WSTF is using progressive finite element analysis methods for test design and for confirmation of composite pressure vessel performance. Using composite finite element analysis models and failure theories tested in the World-Wide Failure Exercise, WSTF is able to estimate the static strength of composite pressure vessels. Additionally, test and evaluation on composites that have been impact damaged is in progress so that models can be developed to estimate damage tolerance and the degradation in static strength.

  12. German boiler and pressure vessel codes and standards: materials, manufacture, testing, equipment, erection and operation

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

    The methods by which the safety objectives on the operation of steam boilers and pressure vessels in Germany can be reached are set out in Technical Rules which are compiled and established in technical committees. Typical applications are described in the Technical Rules. A chart shows how the laws, provisions and Technical Rules for the sections 'steam boiler plant' and 'pressure vessels' are interlinked. This chapter concentrates on legal aspects, materials, manufacture, testing, erection and operation of boilers and pressure vessels in Germany. (U.K.)

  13. A quick guide to API 510 certified pressure vessel inspector syllabus example questions and worked answers

    CERN Document Server

    Matthews, Clifford

    2010-01-01

    The API Individual Certification Programs (ICPs) are well established worldwide in the oil, gas, and petroleum industries. This Quick Guide is unique in providing simple, accessible and well-structured guidance for anyone studying the API 510 Certified Pressure Vessel Inspector syllabus by summarizing and helping them through the syllabus and providing multiple example questions and worked answers.Technical standards are referenced from the API 'body of knowledge' for the examination, i.e. API 510 Pressure vessel inspection, alteration, rerating; API 572 Pressure vessel inspection; API

  14. An introduction to the analysis of multi-cavity prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Silva, M.C.A.T. da.

    1986-01-01

    The present work is a study of multi-cavity prestressed concrete pressure vessels (PCRV) for nuclear reactors. A review is made of the designs, analises and models of multi-cavity concrete pressure vessels. A preliminary evaluation of the NONSAP program for applications in complex three-dimensional structures such as a multi-cavity pressure vessel is also made. A model of a PCRV of a 1000 MW(e) high-temperature gas cooled reactor was selected for a three-dimensional analysis with the NONSAP program. The results obtained are compared with experimental data. (Author) [pt

  15. Blood pressure gradients in cerebral arteries: a clue to pathogenesis of cerebral small vessel disease.

    Science.gov (United States)

    Blanco, Pablo J; Müller, Lucas O; Spence, J David

    2017-09-01

    The role of hypertension in cerebral small vessel disease is poorly understood. At the base of the brain (the 'vascular centrencephalon'), short straight arteries transmit blood pressure directly to small resistance vessels; the cerebral convexity is supplied by long arteries with many branches, resulting in a drop in blood pressure. Hypertensive small vessel disease (lipohyalinosis) causes the classically described lacunar infarctions at the base of the brain; however, periventricular white matter intensities (WMIs) seen on MRI and WMI in subcortical areas over the convexity, which are often also called 'lacunes', probably have different aetiologies. We studied pressure gradients from proximal to distal regions of the cerebral vasculature by mathematical modelling. Blood flow/pressure equations were solved in an Anatomically Detailed Arterial Network (ADAN) model, considering a normotensive and a hypertensive case. Model parameters were suitably modified to account for structural changes in arterial vessels in the hypertensive scenario. Computations predict a marked drop in blood pressure from large and medium-sized cerebral vessels to cerebral peripheral beds. When blood pressure in the brachial artery is 192/113 mm Hg, the pressure in the small arterioles of the posterior parietal artery bed would be only 117/68 mm Hg. In the normotensive case, with blood pressure in the brachial artery of 117/75 mm Hg, the pressure in small parietal arterioles would be only 59/38 mm Hg. These findings have important implications for understanding small vessel disease. The marked pressure gradient across cerebral arteries should be taken into account when evaluating the pathogenesis of small WMIs on MRI. Hypertensive small vessel disease, affecting the arterioles at the base of the brain should be distinguished from small vessel disease in subcortical regions of the convexity and venous disease in the periventricular white matter.

  16. Photoluminescence and optical absorption in neutron-irradiated crystalline quartz

    Energy Technology Data Exchange (ETDEWEB)

    Corazza, A.; Crivelli, B.; Martini, M.; Spinolo, G.; Vedda, A. [Istituto Nazionale Di Fisica Della Materia, Dipartimento di Fisica, Universita di Milano, Via Celoria 16, I-20133 Milano (Italy)

    1996-04-01

    Optical absorption measurements in the 3.5{endash}6.5 eV spectral range and photoluminescence spectra, excited in the 4{endash}8 eV range have been performed on neutron irradiated synthetic crystalline quartz as a function of temperature and of neutron fluence. The Gaussian deconvolution of the radiation-induced absorption spectrum in the 4.5{endash}6 eV region reveals a complex structure: five distinct components, peaking at 4.85, 5.06, 5.35, 5.62, and 5.96 eV are detected. The complexity of the absorption pattern finds a correspondence in photoluminescence spectra excited in the 5 eV region: a detailed analysis of the emission spectra as a function of excitation energy indicates the presence of three emission bands centered at 3.91, 4.23, and 4.46 eV, excited at 5.25, 4.83, and 5.03 eV respectively. Excitation in the 5.62 and 5.96 eV absorption peaks does not produce emission. The features of the 4.23 eV and of the 4.46 eV bands are very similar to those of the {alpha}{sub intrinsic} emission, already well studied in amorphous SiO{sub 2}: this suggests a possible correlation between these bands and the {alpha}{sub intrinsic} center. The 3.91 eV band does not find a correspondence in amorphous SiO{sub 2}, and so the responsible defect appears specifically related to the crystalline structure. The emission spectra excited in the E absorption band ({approx_equal}7.6 eV) present a weak band centered at 4.83 eV: its dependence on neutron irradiation dose suggests the attribution to an intrinsic center different from those responsible for the emission in the 3.8 {endash} 4.5 eV region. {copyright} {ital 1996 The American Physical Society.}

  17. Photoluminescence and optical absorption in neutron-irradiated crystalline quartz

    International Nuclear Information System (INIS)

    Corazza, A.; Crivelli, B.; Martini, M.; Spinolo, G.; Vedda, A.

    1996-01-01

    Optical absorption measurements in the 3.5 endash 6.5 eV spectral range and photoluminescence spectra, excited in the 4 endash 8 eV range have been performed on neutron irradiated synthetic crystalline quartz as a function of temperature and of neutron fluence. The Gaussian deconvolution of the radiation-induced absorption spectrum in the 4.5 endash 6 eV region reveals a complex structure: five distinct components, peaking at 4.85, 5.06, 5.35, 5.62, and 5.96 eV are detected. The complexity of the absorption pattern finds a correspondence in photoluminescence spectra excited in the 5 eV region: a detailed analysis of the emission spectra as a function of excitation energy indicates the presence of three emission bands centered at 3.91, 4.23, and 4.46 eV, excited at 5.25, 4.83, and 5.03 eV respectively. Excitation in the 5.62 and 5.96 eV absorption peaks does not produce emission. The features of the 4.23 eV and of the 4.46 eV bands are very similar to those of the α intrinsic emission, already well studied in amorphous SiO 2 : this suggests a possible correlation between these bands and the α intrinsic center. The 3.91 eV band does not find a correspondence in amorphous SiO 2 , and so the responsible defect appears specifically related to the crystalline structure. The emission spectra excited in the E absorption band (≅7.6 eV) present a weak band centered at 4.83 eV: its dependence on neutron irradiation dose suggests the attribution to an intrinsic center different from those responsible for the emission in the 3.8 endash 4.5 eV region. copyright 1996 The American Physical Society

  18. In Vitro antileishmanial properties of neutron-irradiated meglumine antimoniate

    Directory of Open Access Journals (Sweden)

    Samanta Etel Treiger Borborema

    2005-10-01

    Full Text Available Pentavalent antimony, as meglumine antimoniate (Glucantime® or sodium stibogluconate (Pentostam® , is the main treatment for leishmaniasis, a complex of diseases caused by the protozoan Leishmania, and an endemic and neglected threat in Brazil. Despite over half a century of clinical use, their mechanism of action, toxicity and pharmacokinetic data remain unknown. The analytical methods for determination of antimony in biological systems remain complex and have low sensitivity. Radiotracer studies have a potential in pharmaceutical development. The aim of this study was to obtain a radiotracer for antimony, with suitable physical and biological properties. Meglumine antimoniate was neutron irradiated inside the IEA-R1 nuclear reactor, producing two radioisotopes 122Sb and 124Sb, with high radionuclidic purity and good specific activity. This compound showed the same antileishmanial activity as the native compound. The use of the radiotracers, easily created by neutron irradiation, could be an interesting tool to solve important questions in antimonial pharmacology.Os antimoniais pentavalentes, como o antimoniato de meglumina (Glucantime® ou estibogluconato de sódio (Pentostam® , são o principal tratamento para a leishmaniose, um complexo de doenças causadas pelo protozoário parasita Leishmania, uma doença endêmica e negligenciada no Brasil. Apesar do seu uso clínico por mais de meio século, seu mecanismo de ação, toxicidade e dados de farmacocinética permanecem desconhecidos. Os métodos analíticos para determinação de antimônio em sistemas biológicos são complexos e apresentam baixa sensibilidade. Estudos utilizando radiotraçadores têm papel potencial no desenvolvimento farmacológico. O objetivo deste estudo foi desenvolver um radiotraçador de antimônio, com propriedades físicas e biológicas adequadas. O antimoniato de meglumina foi irradiado por nêutrons no reator nuclear IEA-R1, produzindo dois radioisótopos: 122

  19. Experience of in-service surveillance and monitoring of prestressed concrete pressure vessels for nuclear reactors

    International Nuclear Information System (INIS)

    Irving, J.; Smith, J.R.; Eadie, D.McD.; Hornby, I.W.

    1976-01-01

    Details are given of the statutory requirements for the inspection of prestressed concrete pressure vessels in the United Kingdom, with particular emphasis on the prestressing system. The results of periodic examinations under the Licencing Conditions of the Oldbury and Wylfa vessels are presented and discussed in relation to design expectations and future requirements. Strain, moisture and temperature records obtained from the Oldbury PCPV's over a 10 year period are compared with prediction and new developments in vessel instrumentation are discussed. (author)

  20. Present status of ESNIT (energy selective neutron irradiation test facility) program

    International Nuclear Information System (INIS)

    Noda, K.; Ohno, H.; Sugimoto, M.; Kato, Y.; Matsuo, H.; Watanabe, K.; Kikuchi, T.; Sawai, T.; Usui, T.; Oyama, Y.; Kondo, T.

    1994-01-01

    The present status of technical studies of a high energy neutron irradiation facility, ESNIT (energy selective neutron irradiation test facility), is summarized. Technological survey and feasibility studies of ESNIT have continued since 1988. The results of technical studies of the accelerator, the target and the experimental systems in ESNIT program were reviewed by an International Advisory Committee in February 1993. Recommendations for future R and D on ESNIT program are also summarized in this paper. ((orig.))

  1. Designing of a Fleet-Leader Program for Carbon Composite Overwrapped Pressure Vessels

    Science.gov (United States)

    Murthy, Pappu L.N.; Phoenix, S. Leigh

    2009-01-01

    Composite Overwrapped Pressure Vessels (COPVs) are often used for storing pressurant gases on board spacecraft when mass saving is a prime requirement. Substantial weight savings can be achieved compared to all metallic pressure vessels. For example, on the space shuttle, replacement of all metallic pressure vessels with Kevlar COPVs resulted in a weight savings of about 30 percent. Mass critical space applications such as the Ares and Orion vehicles are currently being planned to use as many COPVs as possible in place of all-metallic pressure vessels to minimize the overall mass of the vehicle. Due to the fact that overwraps are subjected to sustained loads during long periods of a mission, stress rupture failure is a major concern. It is, therefore, important to ascertain the reliability of these vessels by analysis, since it is practically impossible to show by experimental testing the reliability of flight quality vessels. Also, it is a common practice to set aside flight quality vessels as "fleet leaders" in a test program where these vessels are subjected to slightly accelerated operating conditions so that they lead the actual flight vessels both in time and load. The intention of fleet leaders is to provide advanced warning if there is a serious design flaw in the vessels so that a major disaster in the flight vessels can be averted with advance warning. On the other hand, the accelerating conditions must be not so severe as to be prone to false alarms. The primary focus of the present paper is to provide an analytical basis for designing a viable fleet leader program for carbon COPVs. The analysis is based on a stress rupture behavior model incorporating Weibull statistics and power-law sensitivity of life to fiber stress level.

  2. The Combined Effects of Stress Concentration and Tensile Stresses from Autofrettage on the Life of Pressure Vessels

    Science.gov (United States)

    2017-02-01

    PRESSURE VESSELS E. Troiano G.N. Vigilante L.B. Smith J.H. Izzo February 2017 Approved for public...SUBTITLE THE COMBINED EFFECTS OF STRESS CONCENTRATION AND TENSILE STRESSES FROM AUTOFRETTAGE ON THE LIFE OF PRESSURE VESSELS 5a. CONTRACT NUMBER...Approved for public release; distribution is unlimited. 13. SUPPLEMENTARY NOTES 14. ABSTRACT Thick walled pressure vessels are often

  3. High Pressure Composite Overwrapped Pressure Vessel (COPV) Development Tests at Cryogenic Temperatures

    Science.gov (United States)

    Ray, David M.; Greene, Nathanael J.; Revilock, Duane; Sneddon, Kirk; Anselmo, Estelle

    2008-01-01

    Development tests were conducted to evaluate the performance of 2 COPV designs at cryogenic temperatures. This allows for risk reductions for critical components for a Gaseous Helium (GHe) Pressurization Subsystem for an Advanced Propulsion System (APS) which is being proposed for NASA s Constellation project and future exploration missions. It is considered an advanced system since it uses Liquid Methane (LCH4) as the fuel and Liquid Oxygen (LO2) as the oxidizer for the propellant combination mixture. To avoid heating of the propellants to prevent boil-off, the GHe will be stored at subcooled temperatures equivalent to the LO2 temperature. Another advantage of storing GHe at cryogenic temperatures is that more mass of the pressurized GHe can be charged in to a vessel with a smaller volume, hence a smaller COPV, and this creates a significant weight savings versus gases at ambient temperatures. The major challenge of this test plan is to verify that a COPV can safely be used for spacecraft applications to store GHe at a Maximum Operating Pressure (MOP) of 4,500 psig at 140R to 160R (-320 F to -300 F). The COPVs for these tests were provided by ARDE , Inc. who developed a resin system to use at cryogenic conditions and has the capabilities to perform high pressure testing with LN2.

  4. The dynamic relaxation method in the structural analysis of concrete pressure vessels

    International Nuclear Information System (INIS)

    Davidson, I.; Assis Bastos, M.R. de; Camargo, P.B. de.

    1977-01-01

    The dynamic relaxation method, applied to 3 dimensional concrete structures, especially pressure vessels, is demonstrated. It utilizes the finite difference method and allows the growth of cracks to be followed up to the point of vessel rupture. A FORTRAN IV program is developed, which can also be utilized, with the necessary modifications, for other structure calculations [pt

  5. Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

    Science.gov (United States)

    Popelar, Carl F.; Cardinal, Joseph W.

    2014-01-01

    To provide NASA with a suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for the vessels described above, Southwest Research Institute® (SwRI®) was contracted in two phases to obtain relevant material property data from a representative vessel. An initial characterization of the strength, fracture and fatigue crack growth properties was performed in Phase 1. Based on the results and recommendations of Phase 1, a more extensive material property characterization effort was developed in this Phase 2 effort. This Phase 2 characterization included additional strength, fracture and fatigue crack growth of the multilayer vessel and head materials. In addition, some more limited characterization of the welds and heat affected zones (HAZs) were performed. This report

  6. The effect of neutron irradiation on silicon carbide fibers

    International Nuclear Information System (INIS)

    Newsome, G.A.

    1997-01-01

    Nine types of SiC fiber have been exposed to neutron radiation in the Advanced Test Reactor at 250 C for various lengths of time ranging from 83 to 128 days. The effects of these exposures have been initially determined using scanning electron microscopy. The fibers tested were Nicalon trademark CG, Tyranno, Hi-Nicalon trademark, Dow Corning SiC, Carborundum SiC, Textron SCS-6, polymethysilane (PMS) derived SiC from the University of Michigan, and two types of MER SiC fiber. This covers a range of fibers from widely used commercial fibers to developmental fibers. Consistent with previous radiation experiments, Nicalon fiber was severely degraded by the neutron irradiation. Similarly, Tyranno suffered severe degradation. The more advanced fibers which approach the composition and properties of SiC performed well under irradiation. Of these, the Carborundum SiC fiber appeared to perform the best. The Hi-Nicalon and Dow Corning Fibers exhibited good general stability, but also appear to have some surface roughening. The MER fibers and the Textron SCS-6 fibers both had carbon cores which adversely influenced the overall stability of the fibers

  7. Radiation damage in stainless steel under varying temperature neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Naoaki [Kyushu Univ., Kasuga, Fukuoka (Japan). Research Inst. for Applied Mechanics

    1998-03-01

    Microstructural evolution of model alloys of 316SS was examined by neutron irradiation at JMTR under cyclic temperature varying condition. In the case of Fe-16Cr-17Ni, formation of interstitial loops and voids are strongly suppressed by varying the temperature from 473K to 673K. By adding Ti as miner element (0.25wt%), however, abnormal accumulation of vacancies (void swelling of 11%dpa at 0.1dpa) was observed. Theoretical analysis standing on the rate theory of defect clustering and simulation irradiation experiments with heavy ions indicates that the vacancy-rich condition which appears temporally during and after changing the temperature from low to high brings these results. It was also shown that only 1 dpa pre-irradiation at low temperature changes swelling behavior at high temperature above several 10 dpa. The understanding of non-steady-state defect processes under temperature varying irradiation is very important to estimate the radiation damage under fusion environment where short-term and long-term temperature variation is expected. (author)

  8. Neutron-Irradiated Samples as Test Materials for MPEX

    International Nuclear Information System (INIS)

    Ellis, Ronald James; Rapp, Juergen

    2015-01-01

    Plasma Material Interaction (PMI) is a major concern in fusion reactor design and analysis. The Material-Plasma Exposure eXperiment (MPEX) will explore PMI under fusion reactor plasma conditions. Samples with accumulated displacements per atom (DPA) damage produced by fast neutron irradiations in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) will be studied in the MPEX facility. This paper presents assessments of the calculated induced radioactivity and resulting radiation dose rates of a variety of potential fusion reactor plasma-facing materials (such as tungsten). The scientific code packages MCNP and SCALE were used to simulate irradiation of the samples in HFIR including the generation and depletion of nuclides in the material and the subsequent composition, activity levels, gamma radiation fields, and resultant dose rates as a function of cooling time. A challenge of the MPEX project is to minimize the radioactive inventory in the preparation of the samples and the sample dose rates for inclusion in the MPEX facility

  9. Effect of neutron irradiation on select MAX phases

    International Nuclear Information System (INIS)

    Tallman, Darin J.; Hoffman, Elizabeth N.; Caspi, El’ad N.; Garcia-Diaz, Brenda L.; Kohse, Gordon; Sindelar, Robert L.; Barsoum, Michel W.

    2015-01-01

    Herein we report on the effect of neutron irradiation – of up to 0.1 displacements per atom at 360(20) °C or 695(25) °C – on polycrystalline samples of Ti 3 AlC 2 , Ti 2 AlC, Ti 3 SiC 2 and Ti 2 AlN. Rietveld refinement of X-ray diffraction patterns of the irradiated samples showed irradiation-enhanced dissociation into TiC of the Ti 3 AlC 2 and Ti 3 SiC 2 phases, most prominently in the former. Ti 2 AlN also showed an increase in TiN content, as well as Ti 4 AlN 3 after irradiation. In contrast, Ti 2 AlC was quite stable under these irradiation conditions. Dislocation loops are seen to form in Ti 2 AlC and Ti 3 AlC 2 after irradiation at 360(20) °C. The room temperature electrical resistivity of all samples increased by an order of magnitude after irradiation at 360(20) °C, but only by 25% after 695(25) °C, providing evidence for the MAX phases’ dynamic recovery at temperatures as low at 695(25) °C. Based on these preliminary results, it appears that Ti 2 AlC and Ti 3 SiC 2 are the more promising materials for high-temperature nuclear applications

  10. Hyper-thermal neutron irradiation field for neutron capture therapy

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1994-01-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

  11. Effect of neutron irradiation on Mo-Si amorphous alloys

    International Nuclear Information System (INIS)

    Ito, Fumitake; Hasegawa, Masayuki; Suzuki, Kenji; Honda, Toshihisa; Fukunaga, Toshiharu.

    1982-01-01

    The irradiation effects on Mo-Si amorphous alloys were investigated by means of X-ray diffraction and positron annihilation, and their electric resistance at low temperature was measured to examine the superconductivity of the alloys. The specimens of Mo 68 Si 32 and Mo 45 Si 55 were irradiated with the neutron fluence (E > 1 MeV) of about 9 x 10 18 n/cm 2 without temperature control in the Japanese Material Testing Reactor (JMTR). For these irradiated specimens, the X-ray diffraction experiment was performed to examine the irradiation effects on the radial distribution function, and the angular correlation curves for the positron annihilation were also measured. Both experiments showed that there was almost no irradiation effect. However, the width of the superconductive transition measured in Mo 68 Si 32 became extremely narrow due to neutron irradiation, and the transition temperature rose from 6.89 K to 7.03 K. On the other hand, in Mo 45 Si 55 , the width showed a tendency to become somewhat narrow, but the transition temperature shifted to the lower side. (Asami, T.)

  12. Swelling and tensile properties of neutron-irradiated vanadium alloys

    International Nuclear Information System (INIS)

    Loomis, B.A.; Smith, D.L.

    1990-07-01

    Vanadium-base alloys are candidates for use as structural material in magnetic fusion reactors. In comparison to other candidate structural materials (e.g., Type 316 stainless and HT-9 ferritic steels), vanadium-base alloys such as V-15Cr-5Ti and V-20Ti have intrinsically lower long-term neutron activation, neutron irradiation after-heat, biological hazard potential, and neutron-induced helium and hydrogen transmutation rates. Moreover, vanadium-base alloys can withstand a higher surface-heat, flux than steels because of their lower thermal stress factor. In addition to having these favorable neutronic and physical properties, a candidate alloy for use as structural material in a fusion reactor must have dimensional stability, i.e., swelling resistance, and resistance to embrittlement during the reactor lifetime at a level of structural strength commensurate with the reactor operating temperature and structural loads. In this paper, we present experimental results on the swelling and tensile properties of several vanadium-base alloys after irradiation at 420, 520, and 600 degree C to neutron fluences ranging from 0.3 to 1.9 x 10 27 neutrons/m 2 (17 to 114 atom displacements per atom [dpa])

  13. Hot atom reactions in neutron irradiated solid iron group metallocenes

    International Nuclear Information System (INIS)

    Yassine, T.; Blackburn, R.

    1990-01-01

    Investigation of the hot reactions of the recoil metal nuclides in neutron irradiated solid ferrocene, ruthenocene and osmocene shows that the retention decreases from ferrocene to osmocene. This observation is ascribed to the effects of recoil energy and the size of the hot zone. The large isotope effect observed between osmium isotopes and between ruthenium isotopes are explained as resulting from the effect of the Auger ionisation produced by internal conversion which takes place in Os-185 and Ru-97 to a greater degree than in their other isotopes. Irradiation of solid metallocenes diluted with an inert solid (silica) showed that the retention in ruthenocene is high and only slightly less than for the pure case whilst the retention of osmocene is very small and the retention of ferrocene is almost zero. The high retention in solid diluted ruthenocene was tentatively attributed to a combination of primary retention resulting from γ cancellation and uptake of recoil momentum by the solid lattice. (author) 15 refs. 5 tabs. 2 figs

  14. Neutron irradiation of sputtered NbN films

    International Nuclear Information System (INIS)

    Weber, H.W.; Gregshammer, P.; Gray, K.E.; Kampwirth, R.T.

    1989-01-01

    In addition to the excellent high-field superconducting properties of NbN, it is the strain and radiation tolerance, measured in bulk NbN, which makes the material so attractive for large high-field magnets, especially for fusion applications. The authors report neutron irradiation experiments on sputtered NbN films, up to a fluence of 10/sup 23/ m/sup -2/ (E > 0.1 MeV), which prove that NbN films are also extremely radiation-hard high-field superconductors. Both the transition temperatures, T/sub c/, and the normal state resistivities show only small changes with neutron fluence. Concerning the critical current densities, j/sub c/, degradations by as much as 30% are observed at low fields, whereas in an intermediate field range (11-15 T) virtually no change of j/sub c/ and at high fields near 20 T even an increase of j/sub c/ are found. The latter observation is ascribed to a radiation-induced increase of the upper critical field, B/sub c2/, and to the occurence of peak effects near B/sub c2/

  15. Radioactivity of neutron-irradiated cat's-eye chrysoberyls

    Science.gov (United States)

    Tang, S. M.; Tay, T. S.

    1999-04-01

    The recent report of marketing of radioactive chrysoberyl cat's-eyes in South-East Asian markets has led us to use an indirect method to estimate the threat to health these color-enhanced gemstones may pose if worn close to skin. We determined the impurity content of several cat's-eye chrysoberyls from Indian States of Orissa and Kerala using PIXE, and calculated the radioactivity that would be generated from these impurities and the constitutional elements if a chrysoberyl was irradiated by neutrons in a nuclear reactor for color enhancement. Of all the radioactive nuclides that could be created by neutron irradiation, only four ( 46Sc, 51Cr, 54Mn and 59Fe) would not have cooled down within a month after irradiation to the internationally accepted level of specific residual radioactivity of 2 nCi/g. The radioactivity of 46Sc, 51Cr and 59Fe would only fall to this safe limit after 15 months and that of 54Mn could remain above this limit for several years.

  16. Resistivity measurements on the neutron irradiated detector grade silicon materials

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zheng

    1993-11-01

    Resistivity measurements under the condition of no or low electrical field (electrical neutral bulk or ENB condition) have been made on various device configurations on detector grade silicon materials after neutron irradiation. Results of the measurements have shown that the ENB resistivity increases with neutron fluence ({Phi}{sub n}) at low {phi}{sub n} (<10{sup 13} n/cm{sup 2}) and saturates at a value between 300 and 400 k{Omega}-cm at {phi}{sub n} {approximately}10{sup 13} n/cm{sup 2}. Meanwhile, the effective doping concentration N{sub eff} in the space charge region (SCR) obtained from the C-V measurements of fully depleted p{sup +}/n silicon junction detectors has been found to increase nearly linearly with {phi}{sub n} at high fluences ({phi}{sub n} > 10{sup 13} n/cm{sup 2}). The experimental results are explained by the deep levels crossing the Fermi level in the SCR and near perfect compensation in the ENB by all deep levels, resulting in N{sub eff} (SCR) {ne} n or p (free carrier concentrations in the ENB).

  17. Tritium extraction from neutron-irradiated lithium aluminate

    International Nuclear Information System (INIS)

    Garcia H, F.

    1995-01-01

    Lithium aluminate is being strongly considered as a breeder material because of its thermophysical, chemical and mechanical stability at high temperatures and its favorable irradiation behavior. Furthermore, it is compatible with other blanket and structural materials. In this work, the effects of calcination temperature during preparation, extraction temperature and sweep gas composition were observed. Lithium aluminate prepared by four different methods, was neutron irradiated for 30 minutes at a flux of 10 12 -10 13 n/cm 2 s in the TRIGA Mark III reactor at Salazar, Mexico; and the tritium extraction rate was measured. Calcination temperature do not affect the tritium extraction rate. However, using high calcination temperature, gamma lithium aluminate was formed. The tritium extraction at 600 Centigrade degrees was lower than at 800 Centigrade degrees and the tritium amount extracted by distillation of the solid sample was higher. The sweep gas composition showed that tritium extraction was less with Ar plus 0.5 % H 2 that with Ar plus 0.1 % H 2 . This result was contrary to expected, where the tritium extraction rate could be higher when hydrogen is added to the sweep gas. Probably this effect could be attributed to the gas purity. (Author)

  18. Neutron irradiation effect of thermally-sensitized stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Hide, Kouitiro [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) susceptibility of irradiated thermally-sensitized Type 304 Stainless Steels (SSs) was studied as a function of neutron fluence and correlated with mechanical responses of the materials. Neutron irradiation was carried out to neutron fluences up to 1.1 x 10{sup 24} n/m{sup 2} (E > 1MeV) at the light water reactor temperature in the Japan Material Test Reactor. The irradiated specimens were examined by slow strain rate stress corrosion cracking tests in 290degC pure water of 0.2 ppm dissolved oxygen concentration and microhardness measurements. The IGSCC susceptibility of the irradiated specimens increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}. From an attempt to correlate the IGSCC susceptibility with the mechanical properties, an excellent correlation was identified between the susceptibility and microhardness increments at the grain boundary relative to the grain center. While intergranular corrosion rate of thermally sensitized SS increased with neutron fluence up to 1.1 x 10{sup 24} n/m{sup 2}, that of solution annealed SS did not change. The incremental grain boundary hardening and degradation of intergranular corrosion resistance may presumably be the major factors affecting IGSCC performance. (author)

  19. Recovery characteristics of neutron-irradiated V-Ti alloys

    International Nuclear Information System (INIS)

    Leguey, T.; Pareja, R.

    2000-01-01

    The recovery characteristics of neutron-irradiated pure V and V-Ti alloys with 1.0 and 4.5 at.% Ti have been investigated by positron annihilation spectroscopy. Microvoid formation during irradiation at 320 K is produced in pure V and V-1Ti but not in V-4.5Ti. The results are consistent with a model of swelling inhibition induced by vacancy trapping by solute Ti during irradiation. The temperature dependencies of the parameter S in the range 8-300 K indicate a large dislocation bias for vacancies and solute Ti. This dislocation bias prevents the microvoid nucleation in V-4.5Ti, and the microvoid growth in V-1Ti, when vacancies become mobile during post-irradiation annealing treatments. A characteristic increase of the positron lifetime is found during recovery induced by isochronal annealing. It is attributed to a vacancy accumulation into the lattice of Ti oxides precipitated during cooling down, or at their matrix/precipitate interfaces. These precipitates could be produced by the decomposition of metastable phases of Ti oxides formed during post-irradiation annealing above 1000 K

  20. Study of natural diamond detector spectrometric properties under neutron irradiation

    CERN Document Server

    Alekseyev, A B; Kaschuck, Y; Krasilnikov, A; Portnov, D; Tugarinov, S

    2002-01-01

    Natural diamond detector (NDD) performance was studied up to a neutron fluence of 10 sup 1 sup 5 neutron/cm sup 2. The variations of the NDD spectrometric response to incident alpha-particles from sup 2 sup 4 sup 1 Am source after exposure to fast neutron fluences up to 3x10 sup 1 sup 6 n/cm sup 2 were examined. No significant variations up to the level of 10 sup 1 sup 4 n/cm sup 2 were observed. Degradation of charge collection efficiency at higher fluences is reported. No remarkable increase of the NDD leakage current and count rate change had been observed up to a neutron fluence of 3x10 sup 1 sup 6 n/cm sup 2. The charge collection efficiency variations of neutron irradiated diamond spectrometer were studied ex situ under gamma-rays, beta-radiation and visible light excitation. Charge collection efficiency restoration up to 75% level and the NDD performance stabilization by extrinsic low-intensity visible light (550 nm

  1. The measurement for level of marine high-temperature and high-pressure vessels

    International Nuclear Information System (INIS)

    Lin Jie.

    1986-01-01

    The various error factors in measurement for level of marine high-temperature and high-pressure vessels are anslysed. The measuring method of error self compensation and its simplification for land use are shown

  2. Summary of Activities for Health Monitoring of Composite Overwrapped Pressure Vessels Updated January 2014

    Science.gov (United States)

    Skow, Miles G.

    2014-01-01

    This three year project (FY12-14) will design and demonstrate the ability of new Magnetic Stress Gages for the measurement of stresses on the inner diameter of a Composite Overwrapped Pressure Vessel overwrap.

  3. Research progress and recommendations on reactor pressure vessel integrity under hypothetical core melt down accident

    International Nuclear Information System (INIS)

    Yao Yangui; Ning Dong; Wu Zhiwei; Cao Ming; Xie Yongcheng; He Yinbiao; Yao Weida

    2013-01-01

    Background: It is very important to ensure the integrity of the reactor pressure vessel under core melt down accident. The high-temperature creep failure is the main failure mode of the reactor pressure vessel under core melt down accident. Purpose: This paper is to present an overview of research status and progress on high-temperature creep behavior of reactor pressure vessel considering the hypothetical core melt down scenario. Methods: Emphasis is placed on accomplished achievements in creep tests, scale model experiments and numerical simulation, and the domestic newly research productions on high-temperature creep behavior of reactor pressure vessel structure integrity. Conclusions: This paper also discusses the limitations of existing researches and indicates future research directions, such as multi-axis tensile tests, analysis of three-dimensional coupling temperature field, scaled model tests, and so on. (authors)

  4. Selected bibliography on pressure vessels for light-water-cooled power reactors (LWRs)

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1975-01-01

    Abstracts on LWR pressure vessels are arranged in the following categories: general, design, materials technology, fabrication techniques, inspection and testing, and failures. Author, keyword, and KWIC (keyword-in-content) indices are provided. (U.S.)

  5. Mechanical modelling of a structural performance of a pressure vessel submitted to the creep phenomenon

    International Nuclear Information System (INIS)

    Taroco, E.; Feijoo, R.A.; Monteiro, Edson; Freire, J.L.F.; Bevilacqua, L.; Miranda, P.E.V. de; Silveira, T.L. da

    1982-01-01

    A pressure vessel is analized using different mechanical models for the creep phenomenon. The numerical results obtained through these models enable us to recommend on the way verifications of creep damage accumulation is structures should be made. (Author) [pt

  6. Numerical simulation of premixed Hydrogen/air combustion pressure in a spherical vessel

    OpenAIRE

    Guo Han-yu; Tao Gang; Zhang Li-jing

    2016-01-01

    In order to study the development process of hydrogen combustion in a closed vessel, an on-line chemical equilibrium calculator and a numerical simulation method would be used to analysis the combustion pressure and flame front of mixed gas, which based on 20L H2/air explosion experiments in spherical vessel (Crowl and Jo,2009). The results showed that, the turbulent model could reflect the process of combustion, and the error of combustion pressure by simulation is smaller than the Chemical ...

  7. Remote controlled ultrasonic pre-service and in-service inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Mueller, G.

    1990-01-01

    The first mechanised in-service inspection of the reactor pressure vessel on unit one of Eskom's Koeberg nuclear power station has been carried out. Since 1968 a whole range of manipulators to carry out remote controlled ultrasonic inspections of nuclear power station equipment has been developed. The inspection of a reactor pressure vessel using a central mast manipulator is described. 3 figs., 1 ill

  8. Possibility of use small size specimens for toughness evaluation of pressure vessel steels in surveillance programmes

    International Nuclear Information System (INIS)

    Petrequin, P.; Soulat, P.

    1979-01-01

    In order to caracterize fracture toughness evolution of pressure vessel steels under irradiation, two types of small size specimens able to be introduced in surveillance capsules were studied: small (12.5 mm thick) CT specimens and precracked Charpy. Size effects on determination of J1C on irradiated and unirradiated steel are studied using 12.5 and 25 mm thick CT specimens. Kid is measured using precracked Charpy specimens on several pressure vessels irradiated and unirradiated

  9. Evaluations of half-bead weld repair procedures with thick-wall pressure vessels

    International Nuclear Information System (INIS)

    Canonico, D.A.; Whitman, G.D.

    1978-01-01

    The results of research on the evaluation of the half-bead weld repair method for use on nuclear reactor components are reviewed from data obtained on thick-section test pieces and intermediate-size pressure vessels. Material properties, the magnitude of residual stresses and the structural behavior of flawed pressure vessels are being obtained to determine the adequacy of the weld repair method for application in thick-section components

  10. Overview of research trends and problems on Cr-Mo low alloy steels for pressure vessel

    International Nuclear Information System (INIS)

    Chi, Byung Ha; Kim, Jeong Tae

    2000-01-01

    Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding

  11. Positron affinity for precipitates in reactor pressure vessel steels

    Czech Academy of Sciences Publication Activity Database

    Brauer, G.; Puska, M. J.; Šob, Mojmír; Korhonen, T.

    1995-01-01

    Roč. 158, - (1995), s. 149-156 ISSN 0029-5493. [Meeting of the International Group on Radiation Damage Mechanisms in Preassure Vessel Steels /5./. Santa Barbara, 02.05.1994-06.05.1994] Impact factor: 0.134, year: 1995

  12. Comparison of closed-pressurized and open-refluxed vessel ...

    African Journals Online (AJOL)

    Samples of residual fuel oil reference material (SRM 1634c) were mineralized in closed digestion vessels from Milestone Laboratory Systems (MLS) or from PAAR (HPA) or in open-refluxed microwave digestion flasks from Prolabo. The three digestion systems were evaluated in terms of accuracy and precision, reagents ...

  13. Influence of fuel loading on neutron field in WWER-440 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Stacho, M.; Slugen, V.; Farkas, G.; Sojak, S. [Department of Nuclear Physics and Technology, Slovak University of Technology, Ilkovicova 3, 812 19 Bratislava (Slovakia)

    2010-07-01

    One of the limiting factors in terms of nuclear power plant lifetime is reactor pressure vessel neutron load. Neutron embrittlement as the most important ageing effect on the reactor pressure vessel is mainly caused by fast neutron spectra. The work is focused on mapping of neutron fields in the reactor pressure vessel of WWER-440/V-213 reactor using MCNP5 transport code. The calculation of neutron fields was performed using detailed full-core MCNP model of WWER-440 reactor developed at our department. Analysis of fuel loading pattern and burn-up influence on neutron flux density distribution in the reactor pressure vessel was realized. The fuel composition corresponds to fuel cycles of Bohunice and Mochovce nuclear power plants. The goal of this work was to improve the assessment of WWER-440 reactor pressure vessel radiation degradation and following evaluation possibility of its lifetime extension and comparison of neutron flux and neutron spectra in the most loaded place of reactor pressure vessel and surveillance specimen area. (authors)

  14. Evaluation of detectable defect size for inner defect of pressure vessel using laser speckle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyeoung Suk; Seon, Sang Woo; Choi, Tae Ho; Kang, Chan Geun; Na, Man Gyun; Jung, Hyun Chul [Chsoun University, Gwangju (Korea, Republic of)

    2014-04-15

    Pressure vessels are used in various industrial fields. If a defect occurs on the inner or outer surface of a pressure vessel, it may cause a massive accident. A defect on the outer surface can be detected by visual inspection. However, a defect on the inner surface is generally impossible to detect with visual inspection. Nondestructive testing can be used to detect this type of defect. Laser speckle shearing interferometry is one nondestructive testing method that can optically detect a defect; its advantages include noncontact, full field, and real time inspection. This study evaluated the detectable size for an internal defect of a pressure vessel. The material of the pressure vessel was ASTM A53 Gr.B. The internal defect was detected when the pressure vessel was loaded by internal pressure controlled by a pneumatic system. The internal pressure was controlled from 0.2 MPa to 0.6 MPa in increments of 0.2 MPa. The results confirmed that an internal defect with a 25 % defect depth could be detected even at 0.2 MPa pressure variation.

  15. Development of a Numerical Model of Hypervelocity Impact into a Pressurized Composite Overwrapped Pressure Vessel

    Science.gov (United States)

    Garcia, M. A.; Davis, B. A.; Miller, J. E.

    2017-01-01

    As the outlook for space exploration becomes more ambitious and spacecraft travel deeper into space than ever before, it is increasingly important that propulsion systems perform reliably within the space environment. The increased reliability compels designers to increase design margin at the expense of system mass, which contrasts with the need to limit vehicle mass to maximize payload. Such are the factors that motivate the integration of high specific strength composite materials in the construction of pressure vessels commonly referred to as composite overwrapped pressure vessels (COPV). The COPV consists of a metallic liner for the inner shell of the COPV that is stiff, negates fluid permeation and serves as the anchor for composite laminates or filaments, but the liner itself cannot contain the stresses from the pressurant it contains. The compo-site-fiber reinforced polymer (CFRP) is wound around the liner using a combination of hoop (circumferential) and helical orientations. Careful consideration of wrap orientation allows the composite to evenly bear structural loading and creates the COPV's characteristic high strength to weight ratio. As the CFRP overwrap carries most of the stresses induced by pressurization, damage to the overwrap can affect mission duration, mission success and potentially cause loss-of-vehicle/loss-of-crew. For this reason, it is critical to establish a fundamental understanding of the mechanisms involved in the failure of a stressed composite such as that of the COPV. One of the greatest external threats to the integrity of a spacecraft's COPV is an impact from the meteoroid and orbital debris environments (MMOD). These impacts, even from submillimeter particles, generate extremely high stress states in the CFRP that can damage numerous fibers. As a result of this possibility, initial assumptions in survivability analysis for some human-rated NASA space-craft have assumed that any alteration of the vessel due to impact is

  16. Structural considerations in design of lightweight glass-fiber composite pressure vessels

    Science.gov (United States)

    Faddoul, J. R.

    1973-01-01

    The design concepts used for metal-lined glass-fiber composite pressure vessels are described, comparing the structural characteristics of the composite designs with each other and with homogeneous metal pressure vessels. Specific design techniques and available design data are identified. The discussion centers around two distinctly different design concepts, which provide the basis for defining metal lined composite vessels as either (1) thin-metal lined, or (2) glass fiber reinforced (GFR). Both concepts are described and associated development problems are identified and discussed. Relevant fabrication and testing experience from a series of NASA-Lewis Research Center development efforts is presented.

  17. Effect of neutron irradiation on the density of low-energy excitations in vitreous silica. [Neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Terry Lee [Univ. of Illinois, Urbana-Champaign, IL (United States)

    1979-01-01

    Systematic low-temperature measurements of the thermal conductivity, specific heat, dielectric constant, and temperature-dependent ultrasound velocity were made on a single piece of vitreous silica. These measurements were repeated after fast neutron irradiation of the material. It was found that the irradiation produced changes of the same relative magnitude in the low-temperature excess specific heat Cex, the thermal conductivity κ the anomalous temperature dependence of the ultrasound velocity Δv/v. A corresponding change in the temperature dependent dielectric constant was not observed. It is therefore likely that kappa and Δv/v are determined by the same localized excitations responsible for Cex, but the temperature dependence of the dielectric constant may have a different, though possibly related, origin. A consistent account for the measured Cex, κ, and Δv/v of unirradiated silica is given by the tunneling-state model with a single, energy-dependent density of states. Changes in these three properties due to irradiation can be explained by altering only the density of tunneling states incorporated in the model.

  18. Integrated surveillance specimen program for WWER-1000/V-320 reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Kytka, M.; Novosad, P.; Zdarek, J.

    2005-01-01

    Surveillance specimen programs play an important role in reactor pressure vessel lifetime assessment as they should monitor changes in pressure vessel materials, mainly their irradiation embrittlement. Standard surveillance programs in WWER-1000/V-320 reactor pressure vessels have some deficiencies resulting from their design-nonuniformity of neutron field and even within individual specimen sets, large gradient in neutron flux between specimens and containers, lack of neutron monitors in most of containers and no suitable temperature monitors. Moreover, location of surveillance specimens does not assure similar conditions as the beltline region of reactor pressure vessels. Thus, Modified surveillance program for WWER-1000/V-320C type reactors was designed and realized in two units of NPP Temelin, Czech Republic. In this program, large flat type containers are located on inner wall of reactor pressure vessel in the beltline region that assures their practically identical irradiation conditions with critical vessel materials. These containers with inner dimensions of 210 x 300 mm have two layers of specimens; using inserts (10 x 10 x 14 mm) instead of fully Charpy size specimens allows irradiation of materials from several pressure vessels at once in one container. This design advantage has been used for the creation of the Integrated Surveillance Program for several WWER-1000 units-Temelin 1 + 2, Belene (Bulgaria), Rovno 3 + 4, Khmelnick 2, Zaporozhie 6 (Ukraine) and Kalinin 3 (Russia). Irradiation of these archive materials together with the IAEA reference steel JRQ (of ASTM A 533-B type) and reference steel VVER-1000 will allow to compare irradiation embrittlement of these materials and to obtain more reliable and objective results as no reliable predictive formulae exist up to no due to a higher content of nickel in welds. Irradiation of specimens from cladding region will help in the evaluation of resistance of pressure vessels against PTS regimes. (authors)

  19. Reactor pressure vessels safety and reliability - certainty and uncertainty

    International Nuclear Information System (INIS)

    O'Neil, R.

    1977-01-01

    In the paper, it is suggested that the hazard to the population which would result from vessel failure rate of the order of 10 -6 to 10 -7 per vessel year could be acceptable to society on the basis of other natural and man-made risks. The paper considers the problems of demonstrating safety by calculation based on fracture mechanics, and indicates some of the uncertainties, and inconsistencies in the theory, particularly the effect of cracks in locally degraded volumes of material. The phenomenon of crack arrest is considered, and attention is drawn to the uncertainties as indicated at least by some tests. There is need for speedy resolution of this problem. The uncertainties in material properties, heat treatment and residual stresses are considered, and a proposed upper limit for residual defects ('original sin') is proposed. (orig.) [de

  20. Heavy-Section Steel Technology Program intermediate-scale pressure vessel tests

    International Nuclear Information System (INIS)

    Bryan, R.H.; Merkle, J.G.; Smith, G.C.; Whitman, G.D.

    1977-01-01

    The tests of intermediate-size vessels with sharp flaws permitted the comparison of experimentally observed behavior with analytical predictions of the behavior of flawed pressure vessels. Fracture strains estimated by linear elastic fracture mechanics (LEFM) were accurate in the cases in which the flaws resided in regions of high transverse restraint and the fracture toughness was sufficiently low for unstable fracture to occur prior to yielding through the vessel wall. When both of these conditions were not present, unstable fracture did occur, always preceded by stable crack growth; and the cylinders with flaws initially less than halfway through the wall attained gross yield prior to burst. Predictions of failure pressure of the vessels with flawed nozzles, based upon LEFM estimates of failure strain, were very conservative. LEFM calculations of critical load were based upon small-specimen fracture toughness test data. Whenever gross yielding preceded failure, the actual strains achieved were considerably greater than the estimated strains at failure based on LEFM. In such cases the strength of the vessel may be no longer dependent upon plane-strain fracture toughness but upon the capacity of the cracked section to carry the imposed load stably in the plastic range. Stable crack growth, which has not been predictable quantitatively, is an important factor in elastic-plastic analysis of strength. The ability of the flawed vessels to attain gross yield in unflawed sections has important qualitative implications on pressure vessel safety margins. The gross yield condition occurs in light-water-reactor pressure vessels at about 2 x design pressure. The intermediate vessel tests that demonstrated a capacity for exceeding this load confirm that the presumed margin of safety is not diminished by the presence of flaws of substantial size, provided that material properties are adequate

  1. Effect of phase instabilities on the correlation of nickel ion and neutron irradiation swelling in solution annealed 316 stainless steel

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.; Lee, E.H.; Sklad, P.S.

    1979-01-01

    Annealed 316 stainless steel specimens were neutron irradiated to establish steady-state microstructures and then subjected to further high temperature irradiations with 4 MeV Ni ions. It is shown that void growth under neutron irradiation is simulated in ion irradiations carried out at approx. 180 0 C above reactor temperature. However, the precipitate microstructure developed during neutron irradiation is unstable during subsequent ion irradiation. As a result, the relative swelling rates at various reactor temperatures are not simulated correctly

  2. Pressure controlled clamp using shape memory alloy for minimal vessel invasion in blood flow occlusion.

    Science.gov (United States)

    Zhang, Ye; Kanetaka, Hiroyasu; Sano, Yuya; Kano, Mitsuhiro; Kudo, Tadaaki; Sato, Takumi; Shimizu, Yoshinaka

    2013-01-01

    Vessel damage after clamping may affect the success of surgical operations. A new pressure controlled clamp (SMA clamp) was designed using super elastic property of shape memory alloy (SMA) to realize atraumatic vessel occlusion. The ability and biological effect of the SMA clamp to control pressure was investigated in vivo. The loading-displacement curves of the SMA clamps (experimental group) and conventional clamp (control group) by occlusion of pig carotid arteries were evaluated using a clamping-pressure analyzing system. To investigate macroscopically and histologically the vessel damage of the SMA and conventional clamps, pig carotid arteries were stained with Evan's blue and its histological sections were stained with Elastica Massion after clamping for fifteen minutes. Constant value was shown in the loading-displacement curve of SMA clamp. In the control group, damaged area stained with Evan's blue in the vessel wall showed enlargement with the pressure increasing. Less areas in experimental groups are observed than that in the control group. Histological section in the experimental group showed no obvious except a slight compressive damage in the tunica intima. In the control group, vessel wall showed irreversible damages. This experiment indicated that the SMA clamp, which has a unique mechanical property, can be used without vessels damage. This pressure controlled clamp can be a selection in clinical apparatus to improve surgical safety.

  3. Non-Axisymmetric Inflatable Pressure Structure (NAIPS) Concept that Enables Mass Efficient Packageable Pressure Vessels with Sealable Openings

    Science.gov (United States)

    Doggett, William R.; Jones, Thomas C.; Kenner, Winfred S.; Moore, David F.; Watson, Judith J.; Warren, Jerry E.; Makino, Alberto; Yount, Bryan; Selig, Molly; Shariff, Khadijah; hide

    2016-01-01

    Achieving minimal launch volume and mass are always important for space missions, especially for deep space manned missions where the costs required to transport mass to the destination are high and volume in the payload shroud is limited. Pressure vessels are used for many purposes in space missions including habitats, airlocks, and tank farms for fuel or processed resources. A lucrative approach to minimize launch volume is to construct the pressure vessels from soft goods so that they can be compactly packaged for launch and then inflated en route or at the final destination. In addition, there is the potential to reduce system mass because the packaged pressure vessels are inherently robust to launch loads and do not need to be modified from their in-service configuration to survive the launch environment. A novel concept is presented herein, in which sealable openings or hatches into the pressure vessels can also be fabricated from soft goods. To accomplish this, the structural shape is designed to have large regions where one principal stress is near zero. The pressure vessel is also required to have an elongated geometry for applications such as airlocks.

  4. Neutron irradiation effects on magnetic properties of some Heusler alloys

    International Nuclear Information System (INIS)

    Onodera, Hideya; Shinohara, Takeshi; Yamamoto, Hisao; Watanabe, Hiroshi

    1975-01-01

    The neutron irradiation effects were studied with measurements of temperature dependence of magnetization in ordered and disordered Heusler alloys. The irradiation was carried out in JMTR with a total flux of fast neutrons of 10 20 nvt. Fully ordered Cu 2 MnIn, partially ordered Cu 2 MnAl and completely disordered Cu 2 MnSn were prepared with various temperature treatments. The magnetization-temperature curves of each specimen were measured before and after irradiation. In the irradiated Cu 2 MnIn, the disordering by the irradiation gave rise to a decrease of magnetization, and the temperature dependence of magnetization showed that the disordered region contained various regions with different degrees of disorder. For the distribution of the disordered region, the calculation based on the theory of temperature spike by Seitz and Koekler gave a feasible result that a disordered region comprised a central core with a radius of 5.4 A which was completely disordered and a periphery of 3.3 A thickness which was partially disordered. From the magnetization-temperature curves of Cu 2 MnAl, it was considered that the disordered regions induced by the irradiation had different properties from those induced by the heat treatment. The former were the localized and comprised regions corresponding to various degrees of disorder, while the latter spread spatially in a wide range with a certain degree of disorder. The ordering by enhanced diffusion occurred simultaneously to an extent comparable to the disordering, and so it played an important role in the magnetization in the partially disordered Cu 2 MnAl. In the disordered Cu 2 MnSn, however, the ordering effect was very small. It is supposed to be difficult for the A2 structure to transform into the L2 1 structure by the enhanced diffusion. (auth.)

  5. Neutron irradiation behavior of ITER candidate beryllium grades

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B.; Gorokhov, V.A.; Nikolaev, G.N. [A.A.Bochvar All-Russia Scientific Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Melder, R.R.; Ostrovsky, Z.E.

    1998-01-01

    Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 {mu}m, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) {center_dot} 10{sup 21} cm{sup -2} (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T{sub 1} = 130-180degC and T{sub 2} = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 {center_dot} 10{sup 21} cm{sup -2}. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

  6. Study of damages by neutron irradiation in lithium aluminates

    International Nuclear Information System (INIS)

    Palacios G, O.

    1999-01-01

    Lithium aluminates proposed to the production of tritium in fusion nuclear reactors, due to the thermal stability that they present as well as the behavior of the aluminium to the irradiation. As a neutron flux with profile (≅ 14 Mev) of a fusion reactor is not available. A irradiation experiment was designed in order to know the micro and nano structure damages produced by fast and thermal neutrons in two irradiation positions of the fusion nuclear reactor Triga Mark III: CT (Thermal Column) and SIFCA (System of Irradiation Fixed of Capsules). In this work samples of lithium aluminate were characterized by XRD (X-Ray Diffraction), TEM (Transmission Electron Microscopy) and SEM (Scanning Electron Microscopy). Two samples were prepared by two methods: a) coalition method and b) peroxide method. This characterization comprised original and irradiated samples. The irradiated sample amounted to 4 in total: one for each preparation method and one for each irradiation position. The object of this analysis was to correlate with the received neutron dose the damages suffered by the samples with the neutron irradiation during long periods (440 H), in their micro and nano structure aspects; in order to understand the changes as a function of the irradiation zone (with thermal and fast neutron flux) and the preparation methods of the samples and having as an antecedent the irradiation in SIFCA position by short times (2h). The obtained results are referred to the stability of γ -aluminate phase, under given conditions of irradiation and defined nano structure arrangement. They also refer to the proposals of growth mechanism and nucleation of new phases. The error associated with the measurement of neutron dose is also discussed. (Author)

  7. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    Energy Technology Data Exchange (ETDEWEB)

    Stoller, R.E.

    1987-12-01

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.

  8. Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

    International Nuclear Information System (INIS)

    Stoller, R.E.

    1987-12-01

    The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and altered mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs

  9. Further fields of application for prestressed cast iron pressure vessels (PCIV)

    International Nuclear Information System (INIS)

    Guelicher, L.; Schilling, F.E.

    1977-01-01

    The redundancy of the prestressing system of prestressed structures as well as the clear separation of sealing and load-carrying functions of prestressed cast iron pressure vessels offer substantial advantages over conventional welded steel pressure vessels. Because of the temperature resistance of cast iron up to 400 0 C it is possible to build prestressed pressure vessels commercially as hot-working structures. The compressive strength of cast iron, which is 25 times as high as that of concrete allows for a very compact design of the PCIV. Further specific properties of the PCIV like pre-fabrication of the vessel in the production plant - made possible by a structure assembled from segments - short assembly periods at the construction site etc., may open more fields of application. - PCIV as pressurized storage tanks for the emergency shut down system in nuclear power stations. - PCIV as high pressure vessel for the chemical industry. - PCIV as energy storage. - PCIV for light water reactors. - PCIV as burst protection. It is concluded that the application of prestressed cast iron promises to be successful where either structures with large volumes and high pressures and/or temperatures are required or where aspects of safety allow for efficient use of prestressed structures. (Auth.)

  10. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Kori Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Kim, Kwan Hyun; Hong, Joon Wha

    2007-02-15

    This report describes a neutron fluence assessment performed for the Kori Unit 1 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. After Cycle 22 of reactor operation, 2nd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 1 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 23.

  11. Final report for the 2nd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Yonggwang Unit 2 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Yonggwang Unit 2 pressure vessel beltline region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During Cycle 16 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Yonggwang Unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 16.

  12. Final report of the 1st ex-vessel neutron dosimetry installation and evaluations for Yonggwang unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-09-15

    This report describes a neutron fluence assessment performed for the Yonggwang unit 2 pressure vessel beltline region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the beltline region of the pressure vessel. During cycle 15 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Yonggwang unit 2 to provide continuous monitoring of the beltline region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 15.

  13. Final report for the 3rd Ex-Vessel Neutron Dosimetry Installations and Evaluations for Kori Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai (and others)

    2008-03-15

    This report describes a neutron fluence assessment performed for the Kori Unit 1 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. After Cycle 23 of reactor operation, 3rd Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 1 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 24.

  14. Final Report of the 2nd Ex-Vessel Neutron Dosimetry Installation And Evaluations for Yonggwang Unit 1 Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Gong, Un Sik; Choi, Kwon Jae; Chung, Kyoung Ki; Kim, Kwan Hyun; Chang, Jong Hwa; Ha, Jea Ju

    2008-01-15

    This report describes a neutron fluence assessment performed for the Kori Unit 2 pressure vessel belt line region based on the guidance specified in Regulatory Guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During Cycle 21 of reactor operation, an Ex-Vessel Neutron Dosimetry Program was instituted at Kori Unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the Ex-Vessel Neutron Dosimetry Program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-Vessel Neutron Dosimetry has been evaluated at the conclusion of Cycle 21.

  15. Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

    International Nuclear Information System (INIS)

    Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

    2012-01-01

    Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

  16. Modelling creep of pressure vessels with thermal gradients using Theta projection data

    International Nuclear Information System (INIS)

    Law, M.; Payten, W.; Snowden, K.

    2002-01-01

    Pressure vessels are often exposed to through-wall temperature gradients. Thermal stresses occur in addition to pressure stresses. The resulting creep response is calculated using the Theta projection creep algorithm within a finite element code. It was found that the stress and temperature dependence of the creep response may lead to complex stress evolution

  17. Vertical CO2 release experiments from a 1 liter high pressure vessel

    NARCIS (Netherlands)

    Hulsbosch-Dam, C.; Jong, A. de; Zevenbergen, J.; Peeters, R.; Spruijt, M.

    2013-01-01

    High-speed CO2 release experiments from a pressurized vessel have been conducted for a range of storage pressures nozzle diameters. The experiments are conducted to validate and develop models for accidental releases of CO2 during transport. In cooperation with DNV-KEMA lab (Groningen, The

  18. 30 CFR 57.13015 - Inspection of compressed-air receivers and other unfired pressure vessels.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Inspection of compressed-air receivers and...-UNDERGROUND METAL AND NONMETAL MINES Compressed Air and Boilers § 57.13015 Inspection of compressed-air receivers and other unfired pressure vessels. (a) Compressed-air receivers and other unfired pressure...

  19. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

    International Nuclear Information System (INIS)

    Macbeth, R.V.; Trenberth, R.

    1987-12-01

    Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (author)

  20. Ultrasonic stress evaluation through thickness of a stainless steel pressure vessel

    International Nuclear Information System (INIS)

    Javadi, Yashar; Pirzaman, Hamed Salimi; Raeisi, Mohammadreza Hadizadeh; Najafabadi, Mehdi Ahmadi

    2014-01-01

    This paper investigates ultrasonic method in stress measurement through thickness of a pressure vessel. Longitudinal critically refracted (L CR ) waves are employed to measure the welding residual stresses in a vessel constructed from austenitic stainless steel 304L. The acoustoelastic constant is measured through a hydro test to keep the pressure vessel intact. Hoop and axial residual stresses are evaluated by using different frequency range of ultrasonic transducers. The welding processes of vessel shell and caps are simulated by a 3D finite element (FE) model which is validated by hole-drilling method. The residual stresses calculated by FE simulation are then compared with those obtained from the ultrasonic measurement while a good agreement is observed. It is demonstrated that the residual stresses through thickness of the stainless steel pressure vessel can be evaluated by combining FE and L CR method (known as FEL CR method). - Highlights: • The main goal is ultrasonic evaluation of through thickness stresses. • Welding processes of a stainless steel pressure vessel are modelled by FE. • The hole-drilling method is used to validate the FE results. • Residual stresses are measured by four different series of ultrasonic transducers. • The comparison between ultrasonic and FE results show an acceptable agreement