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Sample records for vessel rpv steels

  1. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  2. A phenomenological method of mechanical properties definition of reactor pressure vessels (RPV) steels VVER according to the ball indentation diagram

    International Nuclear Information System (INIS)

    Bakirov, M. B.; Potapov, V.V.; Massoud, J.P.

    2002-01-01

    This work presents specimen-free methods of a standard uniaxial tension diagram construction and RPV (reactor pressure vessel) steels VVER strength properties definition out of a continuous ball indentation diagram. A similarity phenomenon of uniaxial tension strain curves at a hardening area and an area of a ball indentation constitutes the ground of the methods. The methods are developed on the basis of the uniform graphic representation of elasto-plastic strain processes by indentation and tension and with the reception of the unified yield curve at a hardening area. The calculation results on the phenomenological method conducted for a wide range of RPV steels conditions of nuclear reactors have shown a good precision as far as strain curves construction by the uniaxial tension out of the elasto-plastic indentation diagram is concerned. (authors)

  3. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  4. FE-simulation of the viscoplastic behaviour of different RPV steels in the frame of in-vessel melt retentions scenarios

    International Nuclear Information System (INIS)

    Altstadt, E.; Willschuetz, H.G.; Mueller, G.

    2004-01-01

    Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of REactor VEssel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. The geometrical scale of the experiments is 1:10 compared to a common LWR. This paper deals with the experimental, numerical, and metallographical results of the creep failure experiment EC-FOREVER-4, where the American pressure vessel steel SA533B was applied for the lower head. For comparison the results of the experiment EC-FOREVER-3B, build of the French 16MND5 steel, are discussed, too. Emphasis is put on the differences in the viscoplastic behaviour of different heats of the RPV steel. For this purpose, the creep tests in the frame of the LHF/OLHF experiments are reviewed, too. As a hypothesis it is stated that the sulphur content could be responsible for differences in the creep behaviour. (orig.)

  5. 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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPVs) is to thermally anneal them to restore the 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

  7. Initial evaluation of ultrasonic attenuation measurements for estimating fracture toughness of RPV steels

    Energy Technology Data Exchange (ETDEWEB)

    Hiser, A.L. Jr.; Green, R.E. Jr. [Johns Hopkins Univ., Baltimore, MD (United States). Center for Nondestructive Evaluation

    1999-08-01

    Neutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently, there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides initial results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. (orig.)

  8. Irradiation behavior of German PWR RPV steels under operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    May, J.; Hein, H. [AREVA NP Gmbh (Germany); Ganswind, J. [VGB PowerTech e.V. (Germany); Widera, M. [RWE Power AG (Germany)

    2011-07-01

    In 2007, the last standard surveillance capsule of the original RPV (Reactor Pressure Vessel) surveillance programs of the 11 currently operating German PWR has been evaluated. With it the standard irradiation surveillance programs of these plants was completed. In the present paper, irradiation data of these surveillance programs will be presented and a final assessment of the irradiation behavior of the German PWR RPV steels with respect to current standards KTA 3203 and Reg. Guide 1.99 Rev. 2 will be given. Data from two units which are currently under decommissioning will also be included, so that data from all 13 German PWR manufactured by the former Siemens/KWU company (now AREVA NP GmbH) are shown. It will be shown that all surveillance data within the approved area of chemical composition verify the limit curve RT(limit) of the KTA 3203, which is the relevant safety standard for these plants. An analysis of the data shows, that the prediction formulas of Reg. Guide 1.99 Rev. 2 Pos. 1 or from the TTS model tend to overestimate the irradiation behavior of the German PWR RPV steels. Possible reasons for this behavior are discussed. Additionally, the data will be compared to data from the research project CARISMA to demonstrate that these data are representative for the irradiation behavior of the German PWR RPV steels. Since the data of these research projects cover a larger neutron fluence range than the original surveillance data, they offer a future outlook into the irradiation behavior of the German PWR RPV steels under long term conditions. In general, as a consequence of the relatively large and beneficial water gap between core and RPV, especially in all Siemens/KWU 4-loop PWR, the EOL neutron fluence and therefore the irradiation induced changes in mechanical properties of the German PWR RPV materials are rather low. Moreover the irradiation data indicate that the optimized RPV materials specifications that have been applied in particular for the

  9. Reduction of upper shelf energy of highly irradiated RPV steels

    Energy Technology Data Exchange (ETDEWEB)

    Otaka, M.; Osaki, T. [Japan Nuclear Energy Safety Organization (Japan)

    2004-07-01

    It is well known that as the embrittlement due to neutron irradiation of reactor pressure vessel (RPV) steels, there is the tendency of the decrease in Charpy absorbed energy at upper shelf region (USE), in addition to the shift of ductile-brittle transition temperature. Concerning to the regulation of the upper shelf region, no method is provided to evaluate integrity for RPV steels with USE of less than 68J in Japanese codes. Under the circumstance, the reduction tendency of USE using simulated Japanese RPV steels, irradiated by fast neutron up to 1 x 10{sup 24} n/m{sup 2}, E>1 MeV in the OECD Halden test reactor, was investigated to establish the basis of the USE prediction after 60 year plant operation for the integrity assessment of the RPVs. This paper describes the results of an atom probe tomography characterization of irradiated steels. A new form of USE prediction equation was developed based on the atom probe tomography characterization and the Charpy impact test results of the irradiated steels. And, the USE prediction equations have been determined through the regression analysis of the test reactor data combined with Japanese surveillance test data. (orig.)

  10. 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.

  11. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels; RPV-1: un premier reacteur virtuel pour simuler les effets d'irradiation dans les aciers de cuve des reacteurs a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Jumel, St

    2005-01-15

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  12. APT characterization of high nickel RPV steels

    International Nuclear Information System (INIS)

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

    2004-01-01

    Full text: The microstructures of several high nickel content pressure vessel steels have been characterized by atom probe tomography. The purposes of this study were to investigate the influence of high nickel levels on the response to neutron irradiation of high and low copper pressure vessel steels and to establish whether any additional phases were present after neutron irradiation. The nickel levels in these steels were at least twice that typically found in Western pressure vessel steels. Two different types of pressure vessel steels with low and high copper contents were selected for this study. The first set of alloys was low copper (∼0.05% Cu) base (15Ch2NMFAA) and weld (12Ch2N2MAA) materials used in a VVER-1000 reactor. The composition of the lower nickel VVER-1000 base material was Fe- 0.17 wt% C, 0.30% Si, 0.46% Mn, 2.2% Cr, 1.26% Ni, 0.05% Cu, 0.01% S, 0.008% P, 0.10% V and 0.50% Mo. The composition of the higher nickel VVER-1000 weld material was Fe- 0.06 wt % C, 0.33% Si, 0.80% Mn, 1.8% Cr, 1.78% Ni, 0.07% Cu, 0.009% S, 0.005% P, and 0.63% Mo. The VVER-1000 steels were irradiated in the HSSI Program's irradiation facilities at the University of Michigan, Ford Nuclear Reactor at a temperature of 288 o C for 2,137 h at an average flux of 7.08 x 10 11 cm 2 s -1 for a fluence of 5.45 x 10 18 n cm -2 (E >1 MeV) and for 5,340 h at an average flux of 4.33 x 10 11 cm -2 s -1 for a fluence of 8.32 x 10 1 28 n cm -2 (E >1 MeV). Therefore, the total fluence was 1.38 x 10 19 n cm -2 (E >1 MeV). The second type of pressure vessel steel was a high copper (0.20% Cu) weld from the Palisades reactor. The average composition of the Palisades weld was Fe- 0.11 wt% C, 0.18% Si, 1.27% Mn, 0.04% Cr, 1.20% Ni, 0.20% Cu, 0.017% S, 0.014% P, 0.003% V and 0.55% Mn. The Palisades weld, designated weldment 'B' from weld heat 34B009, was irradiated at a temperature of 288 o C and a flux of ∼7 x 10 11 cm -2 s -1 to a fast fluence of 1.4 x 10 19 n cm -2 (E >1 MeV). These three

  13. Thermal embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Corwin, W.R.; Nanstad, R.K.; Alexander, D.J.; Stoller, R.E.; Wang, J.A.; Odette, G.R.

    1995-01-01

    As a result of observations of possible thermal embrittlement from recent studies with welds removed from retired steam generators of the Palisades Nuclear Plant (PNP), an assessment was made of thermal aging of reactor pressure vessel (RPV) steels under nominal reactor operating conditions. Discussions are presented on (1) data from the literature regarding relatively low-temperature thermal embrittlement of RPV steels; (2)relevant data from the US power reactor-embrittlement data base (PR-EDB); and (3)potential mechanisms of thermal embrittlement in low-alloy steels

  14. Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

    International Nuclear Information System (INIS)

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y.

    2014-01-01

    Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10 22 to 3*10 24 n/m 2 depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties), most of the

  15. Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10{sup 22} to 3*10{sup 24} n/m{sup 2} depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties

  16. Long-term aging effects in RPV steel. Final report

    International Nuclear Information System (INIS)

    Bergner, Frank; Ulbricht, Andreas; Wagner, Arne

    2014-01-01

    The BMWi project 1501393 aimed at contributing to the clarification of flux effects and late blooming effects in irradiated RPV steels by means of experimental techniques of sensitivity at the nm scale. The investigation of these effects was focussed on RPV steels, both base metal and weld of German reactors selected according to the objectives of the present project from two previous projects performed at AREVA GmbH. The complementary techniques of small-angle neutron scattering, atom probe tomography and positron annihilation spectroscopy were applied to detect and characterize the irradiation-induced nm-scale defect-solute clusters. A flux effect on the size of the irradiation-induced clusters but no flux effect on both cluster volume fraction and mechanical properties was found. For a low-Cu RPV weld, a late blooming effect was observed, which results in a steep slope of both cluster volume fraction and transition temperature shift after an initial stage of small or no change.

  17. Simulation of creep tests with French or German RPV-steel and investigation of a RPV-support against failure

    International Nuclear Information System (INIS)

    Willschuetz, H.-G.; Altstadt, E.; Sehgal, B.R.; Weiss, F.-P.

    2003-01-01

    Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be considered for a determination of the loadings on the containment. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed in which the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties is performed. For the consideration of the tertiary creep stage and for the evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in three levels: starting with the simulation of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called 'tube-failure-experiments' are modeled: the RUPTHER-14 and the 'MPA-Meppen'-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi5-5 RPV-steels, which are chemically nearly identical. Since these two steels show a similar behavior, it should be allowed to a limited extend to transfer experimental and numerical

  18. Comparison of irradiated and hydrogen implanted German RPV steels using PAS technique

    Energy Technology Data Exchange (ETDEWEB)

    Pecko, Stanislav, E-mail: stanislav.pecko@stuba.sk; Sojak, Stanislav; Slugeň, Vladimír

    2015-12-15

    Highlights: • German RPV steels were originally studied by positron annihilation spectroscopy. • Neutron irradiated and hydrogen ion implanted specimens were studied. • Both irradiation ways caused to increase of defect size. • We determined that the defect size was higher in implanted specimens. - Abstract: Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This spectroscopic method is a really effective tool for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to irradiation. German commercial reactor pressure vessel steels, originally from CARISMA program, were used in our study. The German experimental reactor VAK was selected as the proper irradiation facility in the 1980s. A specimen in as-received state and 2 different irradiated cuts from the same material were measured by PALS and size of defects with their intensity was indentified. Afterwards there was prepared an experiment with concern in simulation of neutron irradiation by hydrogen ion implantation on a linear accelerator with energy of 100 keV. Results are concerning on comparison between defects caused by neutron irradiation and hydrogen implantation. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to hydrogen ions implantation.

  19. Effect of lead factors on the embrittlement of RPV SA-508 cl 3 steel

    Energy Technology Data Exchange (ETDEWEB)

    Kempf, Rodolfo, E-mail: kempf@cnea.gov.ar [CNEA, Unidad Actividad Combustibles Nucleares, División Caracterización, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina); Troiani, Horacio, E-mail: troiani@cab.cnea.gov.ar [Centro Atómico Bariloche (CNEA) e Instituto Balseiro (UNCU), CONICET, Av. Bustillo 9500, CP 8400, Rio Negro (Argentina); Fortis, Ana Maria, E-mail: fortis@cnea.gov.ar [CNEA, Departamento Estructura y Comportamiento, UNSAM, Avda. Gral Paz 1499, C.P.B1650KNA, San Martín, Buenos Aires (Argentina)

    2013-03-15

    This paper presents a project to study the effect of lead factors on the mechanical behaviour of the SA-508 type 3 Reactor Pressure Vessel (RPV) steel used in the reactor under construction Atucha II in Argentina. Charpy-V notch specimens of this steel were irradiated at the RA1 experimental reactor at a temperature of 275 °C with two lead factors (186 and 93). The neutron flux was 3.71 × 10{sup 15} n m{sup −2} s{sup −1} and 1.85 × 10{sup 15} n m{sup −2} s{sup −1} (E > 1 MeV) respectively. In both cases, the fluence was 6.6 × 10{sup 21} n m{sup −2}, which is equivalent to that received by the PHWR Atucha II RPV in 10 years of full power irradiation. The results of Charpy tests revealed significant embrittlement both in the ΔT = 14 °C and ΔT = 21 °C shifts of the ductile–brittle transition temperatures (DBTT) and in the reduction of the maximum energy absorbed. This result shows that the shift of the DBTT with a lead factor of 93 is larger than that obtained with a lead factor of 186. Then, the results of irradiation in experimental reactors (MTR) with high lead factors may not be conservative with respect to the actual RPV embrittlement.

  20. 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.

  1. 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

  2. Statistical evaluation of fracture characteristics of RPV steels in the ductile-brittle transition temperature region

    International Nuclear Information System (INIS)

    Kang, Sung Sik; Chi, Se Hwan; Hong, Jun Hwa

    1998-01-01

    The statistical analysis method was applied to the evaluation of fracture toughness in the ductile-brittle transition temperature region. Because cleavage fracture in steel is of a statistical nature, fracture toughness data or values show a similar statistical trend. Using the three-parameter Weibull distribution, a fracture toughness vs. temperature curve (K-curve) was directly generated from a set of fracture toughness data at a selected temperature. Charpy V-notch impact energy was also used to obtain the K-curve by a K IC -CVN (Charpy V-notch energy) correlation. Furthermore, this method was applied to evaluate the neutron irradiation embrittlement of reactor pressure vessel(RPV) steel. Most of the fracture toughness data were within the 95 percent confidence limits. The prediction of a transition temperature shift by statistical analysis was compared with that from the experimental data. (author)

  3. Mechanical properties and microstructure of long term thermal aged WWER 440 RPV steel

    Energy Technology Data Exchange (ETDEWEB)

    Kolluri, M., E-mail: kolluri@nrg.eu [Nuclear Research & Consultancy Group (NRG), P.O. Box 25, 1755 ZG Petten (Netherlands); Kryukov, A. [Scientific and Engineering Centre for Nuclear and Radiation Safety, 107140 Moscow (Russian Federation); Magielsen, A.J. [Nuclear Research & Consultancy Group (NRG), P.O. Box 25, 1755 ZG Petten (Netherlands); Hähner, P. [European Commission, Joint Research Centre, Directorate G – Nuclear Safety and Security, P.O. Box 2, 1755 ZG Petten (Netherlands); Petrosyan, V. [Armenian Scientific Research Institute for Nuclear Plant Operation (ARMATOM), 0027 Yerevan (Armenia); Sevikyan, G. [Armenian Nuclear Power Plant (ANPP), 0911, Metsamor, Armavir Marz (Armenia); Szaraz, Z. [European Commission, Joint Research Centre, Directorate G – Nuclear Safety and Security, P.O. Box 2, 1755 ZG Petten (Netherlands)

    2017-04-01

    The integrity assessment of the Reactor Pressure Vessel (RPV) is essential for the safe and Long Term Operation (LTO) of a Nuclear Power Plant (NPP). Hardening and embrittlement of RPV caused by neutron irradiation and thermal ageing are main reasons for mechanical properties degradation during the operation of an NPP. The thermal ageing-induced degradation of RPV steels becomes more significant with extended operational lives of NPPs. Consequently, the evaluation of thermal ageing effects is important for the structural integrity assessments required for the lifetime extension of NPPs. As a part of NRG's research programme on Structural Materials for safe-LTO of Light Water Reactor (LWR) RPVs, WWER-440 surveillance specimens, which have been thermal aged for 27 years (∼200,000 h) at 290 °C in a surveillance channel of Armenian-NPP, are investigated. Results from the mechanical and microstructural examination of these thermal aged specimens are presented in this article. The results indicate the absence of significant long term thermal ageing effect of 15Cr2MoV-A steel. No age hardening was detected in aged tensile specimens compared with the as-received condition. There is no difference between the impact properties of as-received and thermal aged weld metals. The upper shelf energy of the aged steel remains the same as for the as-received material at a rather high level of about 120 J. The T{sub 41} value did not change and was found to be about 10 °C. The microstructure of thermal aged weld, consisting carbides, carbonitrides and manganese-silicon inclusions, did not change significantly compared to as-received state. Grain-boundary segregation of phosphorus in long term aged weld is not significant either which has been confirmed by the absence of intergranular fracture increase in the weld. Negligible hardening and embrittlement observed after such long term thermal ageing is attributed to the optimum chemical composition of 15Cr2MoV-A for high

  4. Development of advanced low alloy steel for nuclear RPV

    Energy Technology Data Exchange (ETDEWEB)

    Lee, H. C.; Shin, K. S.; Lee, S. H.; Lee, B. J. [Seoul National Univ., Seoul (Korea)

    2000-04-01

    Low carbon low alloy steels are used in nuclear power plants as pressure vessel, steam generator, etc. Nuclear pressure vessel material requires good combination of strength/ toughness, good weldability and high resistance to neutron irradiation and corrosion fatigue. For SA508III steels, most widely used in the production of nuclear power plant, attaining toughness is more difficult than strength. When taking into account the loss of toughness due to neutron irradiation, attaining as low transition temperature as possible prior to operation is a critical task in the production of nuclear pressure vessels. In the present study, we investigated detrimental microstructural features of SA508III steels to toughness, then alloy design directions to achieve improved mechanical properties were devised. The next step of alloy design was determined based on phase equilibrium thermodynamics and obtained results. Low carbon low alloy steels having low transition temperatures with enough strength and hardenability were developed. Microstructure and mechanical properties of HAZ of SA508III steels and alloy designed steels were investigated. 22 refs., 147 figs., 38 tabs. (Author)

  5. East/west steels for reactor pressure vessels

    International Nuclear Information System (INIS)

    Davies, M.; Kryukov, A.; Nikolaev, Y.; English, C.

    1997-01-01

    The report consist of three parts dealing with comparison of the irradiation behaviour of 'Eastern' and 'Western' steels, mechanisms of irradiation embrittlement and the role of compositional variations on the irradiation sensitivity of pressure vessels. Nickel, copper and phosphorus are the elements rendering the most essential influence on behaviour of pressure vessel steels under irradiation and subsequent thermal annealing. For WWER-440 reactor pressure vessel (RPV) steels in which nickel content does nor exceed 0.3% the main affecting factors are phosphorous and copper. For WWER-1000 RPV welds in which nickel content generally exceed 1.5% the role of nickel in radiation embrittlement is decisive. In 'Western' type steels main influencing elements are nickel and copper. The secondary role of phosphorus in radiation embrittlement of 'Western' steels is caused by lower relative content compared to 'Eastern' steels. The process of how copper, phosphorus and nickel contents affect the irradiation sensitivity of both types of steel seem to be similar. Some distinctions between the observed radiation effects is apparently caused by differences in the irradiation conditions and ratios of the contents of above mentioned elements in both types of steel. For 'Eastern' RPV steels the dependence of the recovery degree of irradiated steels due to postirradiation thermal annealing id obviously dependent on phosphorus contents and the influence of nickel contents on this process is detectable

  6. Vacancy defects in electron irradiated RPV steels studied by positron lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Moser, P; Li, X H [CEA Centre d` Etudes de Grenoble, 38 (France). Dept. de Recherche Fondamentale sur la Matiere Condensee; Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    Specimens of French RPV (reactor pressure vessels) steels at different rates of segregation have been irradiated at 150 and 288 deg C with 3 MeV electrons (irradiation dose: 4*10{sup 19} e-/cm{sup 2}). Vacancy defects are studied by positron lifetime measurements before and after irradiation and at each step of isochronal annealing. After 150 deg C irradiation, a recovery step is observed in both specimens, for annealing treatments in the range 220-370 deg C and is attributed to the dissociation of vacancy-impurity complexes. The size of vacancy clusters never overcome 10 empty atomic volumes. If ``fresh`` dislocations are created just before irradiation, big vacancy clusters could be formed. After 288 deg C irradiation, small vacancy cluster of 4-10 empty atomic volumes are observed. (authors). 3 figs., 7 refs.

  7. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  8. 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)

  9. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  10. Effects of the Microstructure on Segregation behavior of Ni-Cr-Mo High Strength Low Alloy RPV Steel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    SA508 Gr.4N Ni-Cr-Mo low alloy steel has an improved fracture toughness and strength, compared to commercial Mn-Mo-Ni low alloy RPV steel SA508 Gr.3. Higher strength and fracture toughness of low alloy steels could be achieved by adding Ni and Cr. So there are several researches on SA508 Gr.4N low alloy steel for a RPV application. The operation temperature and time of a reactor pressure vessel is more than 300 .deg. C and over 40 years. Therefore, in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel, it requires a resistance of thermal embrittlement in the high temperature range including temper embrittlement resistance. S. Raoul reported that the susceptibility to temper embrittlement was increasing a function of the cooling rate in SA533 steel, which suggests the martensitic microstructures resulting from increased cooling rates are more susceptible to temper embrittlement. However, this result has not been proved yet. So the comparison of temper embrittlement behavior was made between martensitic microstructure and bainitic microstructure with a viewpoint of boundary features in SA508 Gr.4N, which have mixture of tempered bainite/martensite. In this study, we have compared temper embrittlement behaviors of SA508 Gr.4N low alloy steel with changing volume fraction of martensite. The mechanical properties of these low alloy steels) were evaluated after a long-term heat treatment(450 .deg. C, 2000hr. Then, the images of the segregated boundaries were observed and segregation behavior was analyzed by AES. In order to compare the misorientation distributions of model alloys, grain boundary structures were measured with EBSD

  11. LYRA and other projects on RPV steel embrittlement study and mitigation of the AMES network

    International Nuclear Information System (INIS)

    Debarberis, L.; Estorff, U. von; Crutzen, S.; Beers, M.; Stamm, H.; Vries, M.I. de; Tjoa, G.L.

    1998-01-01

    Within the framework of the European Network AMES, Ageing Materials evaluation and Studies, a number of experimental works on RPV materials embrittlement are carried out at the Institute of Advanced Materials (AIM) of the Joint Research Centre (JRC) of the European Commission (EC). The objectives of AMES are mainly the understanding of the property degradation phenomena of RPV western reference steels like JRQ and HSST, eastern RPV steels like 15X2mFA and 15H2X15, and annealing possibilities. In order to conduct a very high quality irradiation rig, LYRA facility, has been designed and developed at the High Flux Reactor (HFR) Petten. An other dedicated rig, named LIMA, has been developed at the HFR Petten in order to irradiate RPV steels, internals and in-core materials under typical BWR/PWR conditions. The samples can be irradiated in pressurised water up to 160 bar, 320 deg. C, and the water chemistry fully controlled. For irradiation of standard or miniaturised LWR related materials samples, another group of well experienced irradiation devices with inert gas or liquid metals environment are employed. These devices are tailored to their various specific applications. This paper is intended to give information about the structure and the objectives of the existing European network AMES, and to present the various AMES main and spin-off projects, including a brief description on he modelling activities related to RPV materials embrittlement. (author)

  12. 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

  13. Microstructural evolution in neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    English, C.A.; Phythian, W.J.

    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. The microstructural evolution in neutron irradiated reactor pressure vessel steels is described. The damage mechanisms are elaborated and techniques for examining the microstructure are suggested. The importance of the initial damage event is analysed, and the microstructural evolution in RPV steels is examined

  14. Fracture toughness prediction for RPV Steels with various degree of embrittlement

    International Nuclear Information System (INIS)

    Margolin, B.; Gulenko, A.; Shvetsova, V.

    2003-01-01

    In the present report, predictions of the temperature dependence of cleavage fracture toughness are performed on the basis of the Master Curve approach and a probabilistic model named now the Prometey model. These predictions are performed for reactor pressure vessel steels in different states, the initial (as-produced), irradiated state with moderate degree of embrittlement and in the highly embrittled state. Calculations of the K IC (T) curves may be performed with both approaches on the basis of fracture toughness test results from pre-cracked Charpy specimens at some (one) temperature. The calculated curves are compared with test results. It is shown that the K IC (T) curves for the initial state calculated with the Master Curve approach and the probabilistic model show good agreement. At the same time, for highly embrittled RPV steel, the K IC (T) curve predicted with the Master Curve approach is not an adequate fit to the experimental data, whereas the agreement of the test results and the K IC (T) curve calculated with the probabilistic model is good. An analysis is performed for a possible variation of the K IC (T) curve shape and the scatter in K IC results. (author)

  15. 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)

  16. JAEA's research on the effects of seawater and radiation on corrosion of Zircaloy and PCV/RPV steels

    International Nuclear Information System (INIS)

    Tsukada, Takashi; Motooka, Takafumi; Nakano, Junichi

    2014-01-01

    In order to implement successfully a lot of work for the extraction of fuel assemblies from spent fuel pool (SFP) and also for the removal of fuel debris from reactor pressure vessel (RPV) and primary containment vessel (PCV) at the Fukushima Daiichi Nuclear Power Station (NPS) of Tokyo Electric Power Co., it is necessary to investigate and to prevent the degradation of structural materials of the fuel assemblies and PCV/RPV which are exposed to the gamma radiation and water containing seawater ingredient, because those factors are influencing and possibly accelerating corrosion of the materials. Therefore, at the Japan Atomic Energy Agency (JAEA), we are carrying out the research related to the corrosion issues which may affect the integrity of fuel assemblies and reactor vessels, i.e. PCV and reactor pressure vessel (RPV), from a viewpoint of the effect of gamma radiation and diluted seawater on corrosion behavior as described in this review. In SFP, hydrazine (N_2H_4) was added to salt-containing water in order to reduce dissolved oxygen (DO). Therefore, deoxygenation behavior by N_2H_4 addition was investigated at the ambient temperature. To evaluate the effects of radiolysis on the initiation of pitting corrosion on Zircaloy-2 in water containing sea salt, the pitting potentials of Zircaloy-2 were evaluated. The experimental results showed that the pitting potential under irradiation was slightly higher than that under conditions in which no radiation was present. Corrosion tests of PCV/RPV steels were conducted in diluted seawater at 50degC under gamma-ray irradiation of dose rates of 4.4 and 0.2 kGy/h. To assess the effect of N_2H_4 as an oxygen scavenger under gamma-ray irradiation in PCV condition, 10 and 100 mg/L N_2H_4 were added to the diluted seawater. When gas phase in test flask was replaced with N_2, corrosion weight loss of the steels decreased remarkably. (author)

  17. Results from Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants - Irradiation Embrittlement of RPV Steels -

    International Nuclear Information System (INIS)

    Abe, Hiroaki; Onizawa, Kunio; Katsuyama, Jinya; Murakami, Kenta; Iwai, Takeo; Iwata, Tadao; Katano, Yoshio; Sekimura, Naoto; Nagai, Yasuyoshi; Toyama, Takeshi; Tamura, Satoshi

    2012-01-01

    As one of the NISA Project on Enhancement of Aging Management and Maintenance in Nuclear Power Plants, we have performed research on the irradiation embrittlement of reactor pressure vessel (RPV) steels, especially focusing on irradiation embrittlement on heat affected zone (HAZ) and on applications of ion beams to deduce fundamental insights irradiation-induced embrittlement. The results obtained from the project are summarized as follows. In order to obtain the technical basis to judge the necessity of surveillance specimens from HAZ, the neutron irradiation program was performed at JRR-3, JAEA. The samples were carefully designed based on the insights from finite element analysis, metallography, 3D atom probe and positron annihilation methods, and were fabricated so as to simulate both heat treatment history and microstructure for typical HAZ from as-fabricated RPV steels which also have variation of impurity levels. The fracture toughness of the unirradiated HAZ specimens was equivalent to or better than that of base metals. Irradiation embrittlement and hardening were roughly identical to those of base metals, while some of the fine-grained HAZ microstructure was susceptible to it. The probabilistic fracture mechanics analysis was applied to the structural integrity assessment taking into account the heterogeneous microstructure as well as susceptibility for irradiation embrittlement of each HAZ microstructure under the variation of welding parameter and PTS condition. It was shown that crack propagation at the fine-grained HAZ, but the discontinuous distribution of the microstructure retards the further propagation. For the precise correlation of irradiation embrittlement of RPV steels for the long term operations, accumulations of high-dose data are required. Ion beam irradiation is one of the solutions for the regime and for mechanism-based descriptions. Another interest of ours was to describe irradiation hardening and embrittlement in terms of

  18. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-01-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain 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 a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  19. Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems

    International Nuclear Information System (INIS)

    Blagoeva, D.T.; Debarberis, L.; Jong, M.; Pierick, P. ten

    2014-01-01

    This paper is illustrating the potential of the well-known low alloyed clean steels, extensively used for the current light water Reactor Pressure Vessels (RPV) steels, for a likely use as a structural material also for the new generation nuclear systems. This option would provide, especially for large components, affordable, easily accessible and a technically more convenient solution in terms of manufacturing and joining techniques. A comprehensive comparison between several sets of surveillance and research data available for a number of RPV clean steels for doses up to 1.5 dpa, and up to 12 dpa for 9%Cr steels, is carried out in order to evaluate radiation stability of the currently used RPV clean steels even at higher doses. Based on the numerous data available, positive preliminary conclusions are drawn regarding the eventual use of clean RPV steels for the massive structural components of the new reactor systems. - Highlights: • Common embrittlement trend between RPV and advanced steels till intermediate doses. • For doses >1.5 dpa, damage rate saturation tendency is observed for RPV steels. • RPV steels might be conveniently utilised also outside their foreseen dose range

  20. Weld residual stresses near the bimetallic interface in clad RPV steel: A comparison between deep-hole drilling and neutron diffraction data

    Energy Technology Data Exchange (ETDEWEB)

    James, M.N., E-mail: mjames@plymouth.ac.uk [School of Marine Science and Engineering, University of Plymouth, Drake Circus, Plymouth (United Kingdom); Department of Mechanical Engineering, Nelson Mandela Metropolitan University, Port Elizabeth (South Africa); Newby, M.; Doubell, P. [Eskom Holdings SOC Ltd, Lower Germiston Road, Rosherville, Johannesburg (South Africa); Hattingh, D.G. [Department of Mechanical Engineering, Nelson Mandela Metropolitan University, Port Elizabeth (South Africa); Serasli, K.; Smith, D.J. [Department of Mechanical Engineering, University of Bristol, Queen' s Building, University Walk, Bristol (United Kingdom)

    2014-07-01

    Highlights: • Identification of residual stress trends across bimetallic interface in stainless clad RPV. • Comparison between deep hole drilling (DHD – stress components in two directions) and neutron diffraction (ND – stress components in three directions). • Results indicate that both techniques can assess the trends in residual stress across the interface. • Neutron diffraction gives more detailed information on transient residual stress peaks. - Abstract: The inner surface of ferritic steel reactor pressure vessels (RPV) is clad with strip welded austenitic stainless steel primarily to increase the long-term corrosion resistance of the ferritic vessel. The strip welding process used in the cladding operation induces significant residual stresses in the clad layer and in the RPV steel substrate, arising both from the thermal cycle and from the very different thermal and mechanical properties of the austenitic clad layer and the ferritic RPV steel. This work measures residual stresses using the deep hole drilling (DHD) and neutron diffraction (ND) techniques and compares residual stress data obtained by the two methods in a stainless clad coupon of A533B Class 2 steel. The results give confidence that both techniques are capable of assessing the trends in residual stresses, and their magnitudes. Significant differences are that the ND data shows greater values of the tensile stress peaks (∼100 MPa) than the DHD data but has a higher systematic error associated with it. The stress peaks are sharper with the ND technique and also differ in spatial position by around 1 mm compared with the DHD technique.

  1. Weld residual stresses near the bimetallic interface in clad RPV steel: A comparison between deep-hole drilling and neutron diffraction data

    International Nuclear Information System (INIS)

    James, M.N.; Newby, M.; Doubell, P.; Hattingh, D.G.; Serasli, K.; Smith, D.J.

    2014-01-01

    Highlights: • Identification of residual stress trends across bimetallic interface in stainless clad RPV. • Comparison between deep hole drilling (DHD – stress components in two directions) and neutron diffraction (ND – stress components in three directions). • Results indicate that both techniques can assess the trends in residual stress across the interface. • Neutron diffraction gives more detailed information on transient residual stress peaks. - Abstract: The inner surface of ferritic steel reactor pressure vessels (RPV) is clad with strip welded austenitic stainless steel primarily to increase the long-term corrosion resistance of the ferritic vessel. The strip welding process used in the cladding operation induces significant residual stresses in the clad layer and in the RPV steel substrate, arising both from the thermal cycle and from the very different thermal and mechanical properties of the austenitic clad layer and the ferritic RPV steel. This work measures residual stresses using the deep hole drilling (DHD) and neutron diffraction (ND) techniques and compares residual stress data obtained by the two methods in a stainless clad coupon of A533B Class 2 steel. The results give confidence that both techniques are capable of assessing the trends in residual stresses, and their magnitudes. Significant differences are that the ND data shows greater values of the tensile stress peaks (∼100 MPa) than the DHD data but has a higher systematic error associated with it. The stress peaks are sharper with the ND technique and also differ in spatial position by around 1 mm compared with the DHD technique

  2. Mechanisms of radiation embrittlement of VVER-1000 RPV steel at irradiation temperatures of (50–400)°C

    Energy Technology Data Exchange (ETDEWEB)

    Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation); National Research Nuclear University “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, Moscow 115409 (Russian Federation); Gurovich, B.A.; Bukina, Z.V.; Frolov, A.S.; Maltsev, D.A.; Krikun, E.V.; Zhurko, D.A.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq. 1, Moscow 123182 (Russian Federation)

    2017-07-15

    This work summarizes and analyzes our recent research results on the effect of irradiation temperature within the range of (50–400)°C on microstructure and properties of 15Kh2NMFAA class 1 steel (VVER-1000 reactor pressure vessel (RPV) base metal). The paper considers the influence of accelerated irradiation with different temperature up to different fluences on the carbide and irradiation-induced phases, radiation defects, yield strength changes and critical brittleness temperature shift (ΔT{sub K}) as well as on changes of the fraction of brittle intergranular fracture and segregation processes in the steel. Low temperature irradiation resulted solely in formation of radiation defects – dislocation loops of high number density, the latter increased with increase in irradiation temperature while their size decreased. In this regard high embrittlement rate observed at low temperature irradiation is only due to the hardening mechanism of radiation embrittlement. Accelerated irradiation at VVER-1000 RPV operating temperature (∼300 °C) caused formation of radiation-induced precipitates and dislocation loops, as well as some increase in phosphorus grain boundary segregation. The observed ΔT{sub K} shift being within the regulatory curve for VVER-1000 RPV base metal is due to both hardening and non-hardening mechanisms of radiation embrittlement. Irradiation at elevated temperature caused more intense phosphorus grain boundary segregation, but no formation of radiation-induced precipitates or dislocation loops in contrast to irradiation at 300 °C. Carbide transformations observed only after irradiation at 400 °C caused increase in yield strength and, along with a contribution of the non-hardening mechanism, resulted in the lowest ΔT{sub K} shift in the studied range of irradiation temperature and fluence. - Highlights: •Structural elements in RPV steel are studied at different irradiation temperatures. •Highest number density dislocation loops are

  3. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  4. Effects of the phase fractions on the carbide morphologies, Charpy and tensile properties in SA508 Gr.4N High Strength Low Alloy RPV Steel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Wee, Dang Moon [KAIST, Daejeon (Korea, Republic of); Kim, Min Chul; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    To improve the strength and toughness of RPV (reactor pressure vessel) steels for nuclear power plants, an effective way is the change of material specification from tempered bainitic SA508 Gr.3 Mn-Mo-Ni low alloy steel into tempered martensitic/bainitic SA508 Gr.4N Ni-Cr-Mo low alloy steel. It is known that the phase fractions of martensitic/bainitic steels are very sensitive to the austenitizing cooling rates. Kim reported that there are large differences of austenitizing cooling rates between the surface and the center locations in RPV due to its thickness of 250mm. Hence, the martensite/bainite fractions would be changed in different locations, and it would affect the microstructure and mechanical properties in Ni-Cr-Mo low alloy steel. These results may lead to inhomogeneous characteristics after austenitizing. Therefore, it is necessary to evaluate the changes of microstructure and mechanical properties with varying phase fractions in Ni-Cr-Mo low alloy steel. In this study, the effects of martensite/bainite fractions on microstructure and mechanical properties in Ni-Cr-Mo low alloy steel were examined. The changes in phase fractions of Ni-Cr-Mo low alloy steel with different cooling rates were analyzed, and then the phase fractions were correlated with its microstructural observation and mechanical properties

  5. Validation of software components for the prediction of irradiation-induced damage of RPV steel

    International Nuclear Information System (INIS)

    Bergner, Frank; Birkenheuer, Uwe; Ulbricht, Andreas

    2010-04-01

    The modelling of irradiation-induced damage of RPV steels from primary cascades up to the change of mechanical properties bridging length scales from the atomic level up to the macro-scale and time scales up to years contributes essentially to an improved understanding of the phenomenon of neutron embrittlement. In future modelling may become a constituent of the procedure to evaluate RPV safety. The selected two-step approach is based upon the coupling of a rate-theory module aimed at simulating the evolution of the size distribution of defect-solute clusters with a hardening module aimed at predicting the yield stress increase. The scope of the investigation consists in the development and validation of corresponding numerical tools. In order to validate these tools, the output of representative simulations is compared with results from small-angle neutron scattering experiments and tensile tests performed for neutron-irradiated RPV steels. Using the developed rate-theory module it is possible to simulate the evolution of size, concentration and composition of mixed Cu-vacancy clusters over the relevant ranges of size up to 10.000 atoms and time up to tens of years. The connection between the rate-theory model and hardening is based upon both the mean spacing and the strength of obstacles for dislocation glide. As a result of the validation procedure of the numerical tools, we have found that essential trends of the irradiation-induced yield stress increase of Cu-bearing and low-Cu RPV steels are displayed correctly. First ideas on how to take into account the effect of Ni on both cluster evolution and hardening are worked out.

  6. Progress in RPV-examination of the Chooz-A vessel (and the French procedures, its new developments (MIS5))

    Energy Technology Data Exchange (ETDEWEB)

    Samman, J; Martin, E; Lacroix, R [Electricite de France (EDF), 93 - Saint-Denis (France). Groupe des Labs.

    1988-12-31

    This document deals with the French Chooz-A reactor. It describes the method used for in-service inspection of Reactor Pressure Vessels (RPV). The ultrasonic testing procedure is described, showing its advantages and limitations. The supplementary ultrasonic examination is also described, as well as the validation of underclad cracks detection and sizing. Historical data is also provided. (TEC).

  7. 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

  8. RUPTHER - an original experimental approach for creep failure study of RPV steel

    International Nuclear Information System (INIS)

    Sainte Catherine, C.; Mongabure, Ph.; Cotoni, V.; Nicolas, L.; Devos, J.

    1998-01-01

    Rupter (Rupture Under Thermal Conditions) experiment is designed in order to get validated models for the degradation of RPV (Reactor Pressure Vessel) bottom head in case of a severe accident with corium flow. A simple experimental testing device has been designed in order to perform realistic thermo-mechanical loading on a cylinder. It is externally heated in its central part by induction (max. 1300 deg C) giving an axial thermal gradient. The cylinder is then mechanically loaded by internal pressure (max. 100 bars) until failure occurrence. (authors)

  9. Comparison of the segregation behavior between tempered martensite and tempered bainite in Ni-Cr-Mo high strength low alloy RPV steel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Kim, Min Chul; Kim, Hyung Jun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    SA508 Gr.4N Ni-Cr-Mo low alloy steel has an superior fracture toughness and strength, compared to commercial Mn-Mo-Ni low alloy RPV steel SA508 Gr.3. Higher strength and fracture toughness of low alloy steels could be obtained by adding Ni and Cr. So several were performed on researches on SA508 Gr.4N low alloy steel for a RPV application. The operation temperature and term of a reactor pressure vessel is more than 300 .deg. C and over 40 years. Therefore, in order to apply the SA508 Gr.4N low alloy steel for a reactor pressure vessel, the resistance of thermal embrittlement in the high temperature range including temper embrittlement is required. S. Raoul reported that the susceptibility to temper embrittlement was increasing a function of the cooling rate in SA533 steel, which suggests the martensitic microstructures resulting from increased cooling rates are more susceptible to temper embrittlement. However, this result has not been proved yet. So the comparison of temper embrittlement behavior was made between martensitic microstructure and bainitic microstructure with a viewpoint of boundary features in SA508 Gr.4N, which have mixture of tempered bainite/martensite. We have compared temper embrittlement behaviors of SA508 Gr.4N low alloy steel with changing volume fraction of martensite. The mechanical properties of these low alloy steels were evaluated after a long-term heat treatment. Then, the the segregated boundaries were observed and segregation behavior was analyzed by AES. In order to compare the misorientation distributions of model alloys, grain boundary structures were measured with EBSD

  10. Status Summary of FY16 Atom Probe Tomography Studies on UCSB ATR-2 Irradiated RPV Steels

    Energy Technology Data Exchange (ETDEWEB)

    Wells, Peter [Idaho National Lab. (INL), Idaho Falls, ID (United States); Odette, G. Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-05-01

    The University of California Santa Barbara-2 RPV Steel Irradiation experiment was awarded in 2010 by the Nuclear Science User Facility (formerly ATR NSUF) through a competitive peer review proposal process. The experiment involved irradiation of nearly 1300 samples distributed over 13 capsules. The major objective of this experiment was to better understand embrittlement behavior of reactor pressure steels at doses beyond which available data exists yet may be achieved if reactor operating licenses are extended beyond 60 years. The experiment was instrumented during irradiation and active temperature control was used to maintain the temperature at the design temperature. Six samples were selected from a large matrix of materials to perform atom probe tomography (APT) to look at formation of high dose phases. The nature and formation behavior of these phases is discussed.

  11. Comparison of different experimental and analytical measures of the thermal annealing response of neutron-irradiated RPV steels

    International Nuclear Information System (INIS)

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

    1997-01-01

    The thermal annealing response of several materials as indicated by Charpy transition temperature (TT) and upper-shelf energy (USE), crack initiation toughness, K Jc , predictive models, and automated-ball indentation (ABI) testing are compared. The materials investigated are representative reactor pressure vessel (RPV) steels (several welds and a plate) that were irradiated for other tasks of the Heavy-Section Steel Irradiation (HSSI) Program and are relatively well characterized in the unirradiated and irradiated conditions. They have been annealed at two temperatures, 343 and 454 C (650 and 850 F) for varying lengths of time. The correlation of the Charpy response and the fracture toughness, ABI, and the response predicted by the annealing model of Eason et al. for these conditions and materials appears to be reasonable. The USE after annealing at the temperature of 454 C appears to recover at a faster rate than the TT, and even over-recovers (i.e., the recovered USE exceeds that of the unirradiated material)

  12. Influence of segregations and hydrogen flakes on the mechanical properties of forged RPV steels

    International Nuclear Information System (INIS)

    Eiselt, C.C.; May, J.; Hein, H.

    2013-01-01

    In the frame of relevant 1970s/80s German research programs (e.g. FKS research program on component safety and others), many investigations on large forgings manufactured from Reactor Pressure Vessel (RPV) materials such as 20 MnMoNi 5 5 and 22 NiMoCr 3 7 have been performed. Lately, after ultrasonic testing hydrogen flakes in connection with segregation zones have been observed in a few RPV forgings. The earlier R and D programs contained a number of special heats, which covered a defined defect state (lower bound heats) with relevance to the recent observations of numerous UT indications in RPV forgings of two PWRs. Therefore, the results of these former research programs were now reviewed. The studies included an evaluation of the effects of macro/micro segregations as well as hydrogen flakes on the mechanical properties. As part of the mechanical technological experiments Charpy impact tests in different orientations (e.g. L-T, T-L and S-T) together with fracture mechanics and large scale tensile tests were carried out in segregated and non segregated material zones. In this context the letters L,T,S indicate the longitudinal, transversal and short transverse (thickness) direction with respect to rolling direction of the forging axis. The first letter indicates the direction of the principal stress, while the second letter stands for the crack propagation direction [1]. Furthermore the irradiation behavior of segregated material regions was analyzed and compared to non segregated material regions. Key results of these analyses indicate that in most cases upper shelf levels are lowered in segregated material parts compared to non segregated areas. In addition the segregations cause a larger scattering of impact energies. A high hydrogen content in combination with segregations has overall detrimental effects on the mechanical properties. However, there seems to be no specific segregation influence on the materials' irradiation reaction.

  13. Account of the effect of nuclear collision cascades in model of radiation damage of RPV steels

    International Nuclear Information System (INIS)

    Kevorkyan, Yu.R.; Nikolaev, Yu.A.

    1997-01-01

    A kinetic model is proposed for describing the effect of collision cascades in model of radiation damage of reactor pressure vessel steels. This is a closed system of equations which can be solved only by numerical methods in general case

  14. Correlation between microstructural features and mechanical properties of irradiated LONGLIFE RPV steels

    International Nuclear Information System (INIS)

    Serrano, M.; Hermandez-Mayoral, E.; Bergner, F.; Viehrig, H.W.; Altstadt, E.; Radiguet, B.; Lim, J.H.; Grovenor, C.R.M.; Meslin, E.; Van Renterghem, W.; Chaouadi, R.; Ortner, S.; Hein, H.; Gillemot, F.; Todeschini, P.; Planman, T.; Wilford, K.; Kocik, J.; Brumovsky, M.; Ruoden, J.

    2015-01-01

    The possibility of extending the operational life of reactor pressure vessels (RPV) up to 80 years presents the problem of the availability of materials irradiated at high neutron fluence and low neutron flux. The ability of the existing trend curves to predict high fluence embrittlement is a question of debate, and a critical analysis of these curves should be based on a consistent microstructural examination of irradiated materials. Within the LONGLIFE 7FWP, neutron irradiated RPV materials, relevant for long term operation, some of them coming from surveillance programs, have been characterized by means of a combination of microstructural techniques (APT, SANS, TEM) and mechanical tests (hardness, tensile, impact and fracture toughness). In this paper the analysis of the links between microstructural features (solute nano-clusters, dislocation loops and voids) and hardening and embrittlement measurements by mechanical testing, is presented. Current hardening models, based on the contribution of precipitates, or nano-clusters, seem to underestimate irradiation hardening for very high fluences, mainly when matrix damage (dislocation loops) is observed. Regarding chemical composition effects, the predominant role of Ni and the synergism between Ni-Mn and Si are also identified. Low-Cu alloys show a threshold value of radiation induced features to produce an effect on mechanical properties which calls for further in-depth analyses. (authors)

  15. Device for investigating subcritical crack growth of RPV steel specimens under BWR conditions

    International Nuclear Information System (INIS)

    Anders, D.; Ahlf, J.

    1983-01-01

    An experiment is being prepared to investigate the subcritical crack growth of RPV steel specimens under cyclic load and under the environmental conditions of a BWR with regard to primary water and irradiation. The experiment will be carried out in the VAK reactor Kahl which is a boiling water reactor operating at 71 bar, 286 0 C and generating 16 MW/sub e/. The experimental setup is composed of an open frame to which a string consisting of five compact tension speciments (40 mm thickness) and connecting links is fixed. The specimen chain is set under cyclic load by a pneumatically actuated bellows unit which is attached to the frame top. Specimen strain and crack opening are measured by linear differential transformers; for temperature distribution measurements in the specimens thermocouples are applied

  16. Characterization by notched and precracked Charpy tests of the in-service degradation of RPV steel fracture toughness

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.

    1997-01-01

    The current engineering and regulatory practice to estimate fracture toughness safety margins for nuclear reactor pressure vessels (RPVs) relies heavily on the CVN impact test. Techniques to estimate in-service toughness degradation directly using a variety of precracked specimens are under development worldwide. Emphasis is on their miniaturization. In the nuclear context, it is essential to address many issues such as representativity of the surveillance programs with respect to the vessel in terms of materials and environment, transferability of test results to the structure (constraint and size effects), lower bound toughness certification, creadibility relative to trends of exising databases. An enhanced RPV surveillance strategy in under development in Belgium. It combines state-of-the-art micromechanical and damage modelling to the evaluation of CVN load-deflection signals, tensile stress-strain curves and slow-bend tests of reconstituted precracked Charpy specimens. A probabilistic micromechanical model has been established for static and dynamic transgranular cleavage initiation fracture toughness in the ductile-brittle transition temperature range. This model allows to project toughness bounds for any steel embrittlement condition from the corresponding CVN and static tensile properties, using a single scaling factor defined by imposing agreement with toughness tests in a single condition. The outstanding finding incorporated by this toughness transfer model is that the microcleavage fracture stress is affected by temperature in the ductile-brittle transition and that this influence is strongly correlated to the flow stress: this explains the shape of the K{sub Ic}n K{sub Id} temperature curves as well as the actual magnitude of the strain rate and irradiation effects. Furthermore, CVN crack arrest loads and fracture appearance are also taken advantage of in order to estimate K{sub Ia} degradation. Finally, the CVN-tensile load-temperature diagram

  17. Characterization by notched and precracked Charpy tests of the in-service degradation of RPV steel fracture toughness

    International Nuclear Information System (INIS)

    Fabry, A.

    1997-01-01

    The current engineering and regulatory practice to estimate fracture toughness safety margins for nuclear reactor pressure vessels (RPVs) relies heavily on the CVN impact test. Techniques to estimate in-service toughness degradation directly using a variety of precracked specimens are under development worldwide. Emphasis is on their miniaturization. In the nuclear context, it is essential to address many issues such as representativity of the surveillance programs with respect to the vessel in terms of materials and environment, transferability of test results to the structure (constraint and size effects), lower bound toughness certification, creadibility relative to trends of exising databases. An enhanced RPV surveillance strategy in under development in Belgium. It combines state-of-the-art micromechanical and damage modelling to the evaluation of CVN load-deflection signals, tensile stress-strain curves and slow-bend tests of reconstituted precracked Charpy specimens. A probabilistic micromechanical model has been established for static and dynamic transgranular cleavage initiation fracture toughness in the ductile-brittle transition temperature range. This model allows to project toughness bounds for any steel embrittlement condition from the corresponding CVN and static tensile properties, using a single scaling factor defined by imposing agreement with toughness tests in a single condition. The outstanding finding incorporated by this toughness transfer model is that the microcleavage fracture stress is affected by temperature in the ductile-brittle transition and that this influence is strongly correlated to the flow stress: this explains the shape of the K Ic n K Id temperature curves as well as the actual magnitude of the strain rate and irradiation effects. Furthermore, CVN crack arrest loads and fracture appearance are also taken advantage of in order to estimate K Ia degradation. Finally, the CVN-tensile load-temperature diagram provides substantial

  18. Nondestructive characterization of embrittlement in reactor pressure vessel steels -- A feasibility study

    International Nuclear Information System (INIS)

    McHenry, H.I.; Alers, G.A.

    1998-01-01

    The Nuclear Regulatory Commission recently initiated a study by NIST to assess the feasibility of using physical-property measurements for evaluating radiation embrittlement in reactor pressure vessel (RPV) steels. Ultrasonic and magnetic measurements provide the most promising approaches for nondestructive characterization of RPV steels because elastic waves and magnetic fields can sense the microstructural changes that embrittle materials. The microstructural changes of particular interest are copper precipitation hardening, which is the likely cause of radiation embrittlement in RPV steels, and the loss of dislocation mobility that is an attribute of the ductile-to-brittle transition. Measurements were made on a 1% copper steel, ASTM grade A710, in the annealed, peak-aged and overaged conditions, and on an RPV steel, ASTM grade A533B. Nonlinear ultrasonic and micromagnetic techniques were the most promising measures of precipitation hardening. Ultrasonic velocity measurements and the magnetic properties associated with hysteresis-loop measurements were not particularly sensitive to either precipitation hardening or the ductile-to-brittle transition. Measurements of internal friction using trapped ultrasonic resonance modes detected energy losses due to the motion of pinned dislocations; however, the ultrasonic attenuation associated with these measurements was small compared to the attenuation caused by beam spreading that would occur in conventional ultrasonic testing of RPVs

  19. Development of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed

  20. Development of PWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed.

  1. A micromechanical interpretation of the temperature dependence of Beremin model parameters for French RPV steel

    International Nuclear Information System (INIS)

    Mathieu, Jean-Philippe; Inal, Karim; Berveiller, Sophie; Diard, Olivier

    2010-01-01

    Local approach to brittle fracture for low-alloyed steels is discussed in this paper. A bibliographical introduction intends to highlight general trends and consensual points of the topic and evokes debatable aspects. French RPV steel 16MND5 (equ. ASTM A508 Cl.3), is then used as a model material to study the influence of temperature on brittle fracture. A micromechanical modelling of brittle fracture at the elementary volume scale already used in previous work is then recalled. It involves a multiscale modelling of microstructural plasticity which has been tuned on experimental inter-phase and inter-granular stresses heterogeneities measurements. Fracture probability of the elementary volume can then be computed using a randomly attributed defect size distribution based on realistic carbides repartition. This defect distribution is then deterministically correlated to stress heterogeneities simulated within the microstructure using a weakest-link hypothesis on the elementary volume, which results in a deterministic stress to fracture. Repeating the process allows to compute Weibull parameters on the elementary volume. This tool is then used to investigate the physical mechanisms that could explain the already experimentally observed temperature dependence of Beremin's parameter for 16MND5 steel. It is showed that, assuming that the hypothesis made in this work about cleavage micro-mechanisms are correct, effective equivalent surface energy (i.e. surface energy plus plastically dissipated energy when blunting the crack tip) for propagating a crack has to be temperature dependent to explain Beremin's parameters temperature evolution.

  2. Corrosion fatigue crack growth behaviour of low-alloy RPV steels at different temperatures and loading frequencies under BWR/NWC environment

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.

    2004-01-01

    The strain-induced corrosion cracking or low-frequency corrosion fatigue (LFCF) crack growth behaviour of different reactor pressure vessel (RPV) steels and of a RPV weld filler/weld heat-affected zone (HAZ) material were characterized under simulated transient boiling water reactor/normal water chemistry conditions by cyclic fatigue tests with pre-cracked fracture mechanics specimens. The experiments were performed in oxygenated high-temperature water at temperatures of either 288, 250, 200, or 150 deg. C. Modern high-temperature water loops, on-line crack growth monitoring (DCPD) and fractographic analysis by SEM were used to quantify the cracking response. Under low-flow and highly oxidising conditions (ECP > 0 mV SHE , O 2 = 0.4 ppm) the cycle-based LFCF crack growth rates (CGR) Δa/ΔN increased with decreasing loading frequency and increasing temperature with a maximum/plateau at/above 250 deg. C. Sustained environmentally-assisted crack growth could be maintained down to low frequencies of 10 -5 Hz. The LFCF CGR of low- and high-sulphur steels and of the weld filler/HAZ material were comparable over a wide range of loading conditions and conservatively covered by the 'high-sulphur line' of the General Electric-model. The 'ASME XI wet fatigue CGR curves' could be significantly exceeded in all materials by cyclic fatigue loading at low frequencies ( -2 Hz) at high and low load ratios R. (authors)

  3. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 {sup o}C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV

  4. PSI contribution to the CASTOC round robin on EAC of low-alloy RPV steels under BWR conditions

    International Nuclear Information System (INIS)

    Seifert, H.P.; Ritter, S.

    2001-08-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack growth (EAC) behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state BWR power operation conditions by 6 European laboratories. The present report contains a summary of the PSI contribution to the Working Package 1 (WP1) of this project. WP1 is an interlaboratory round robin EAC test in simulated BWR/NWC environment under cyclic and static loading conditions. The round robin shall demonstrate the applicability of the used advanced test technique and establishes the technical basis for the decision of test conditions in the other working packages. In the first part of the report, the PSI testing facility/measurement instruments and the applied test and evaluation procedure are discussed in detail. In the second part, the exact test conditions and test results with detailed post-test fractographical evaluation in the SEM are presented. The test results are compared with other PSI results, literature data and nuclear codes. Stable and stationary test conditions within the specified range could be achieved in the PSI test during the whole conditioning and experimental phase. The cyclic crack growth rate results agree well with recent PSI results at a higher dissolved oxygen content of 8 ppm and are slightly below the 'high-sulphur line' of the PLEDGE-model. The crack growth rates are significantly above the ASME XI 'wet' curve. Compared to fatigue crack growth rates in air under otherwise identical test conditions, the effect of the high-temperature water environment resulted in an acceleration of crack growth by a factor of 150-250 under these low-cyclic loading conditions. The test results at constant load confirm the extremely low susceptibility to SCC crack growth under static load at 288 o C observed in tests at MPA, PSI and in a European Round Robin. They agree well with the RPV operating experience

  5. Application of Bimodal Master Curve Approach on KSNP RPV steel SA508 Gr. 3

    International Nuclear Information System (INIS)

    Kim, Jongmin; Kim, Minchul; Choi, Kwonjae; Lee, Bongsang

    2014-01-01

    In this paper, the standard MC approach and BMC are applied to the forging material of the KSNP RPV steel SA508 Gr. 3. A series of fracture toughness tests were conducted in the DBTT transition region, and fracture toughness specimens were extracted from four regions, i.e., the surface, 1/8T, 1/4T and 1/2T. Deterministic material inhomogeneity was reviewed through a conventional MC approach and the random inhomogeneity was evaluated by BMC. In the present paper, four regions, surface, 1/8T, 1/4T and 1/2T, were considered for the fracture toughness specimens of KSNP (Korean Standard Nuclear Plant) SA508 Gr. 3 steel to provide deterministic material inhomogeneity and review the applicability of BMC. T0 determined by a conventional MC has a low value owing to the higher quenching rate at the surface as expected. However, more than about 15% of the KJC values lay above the 95% probability curves indexed with the standard MC T0 at the surface and 1/8T, which implies the existence of inhomogeneity in the material. To review the applicability of the BMC method, the deterministic inhomogeneity owing to the extraction location and quenching rate is treated as random inhomogeneity. Although the lower bound and upper bound curve of the BMC covered more KJC values than that of the conventional MC, there is no significant relationship between the BMC analysis lines and measured KJC values in the higher toughness distribution, and BMC and MC provide almost the same T0 values. Therefore, the standard MC evaluation method for this material is appropriate even though the standard MC has a narrow upper/lower bound curve range from the RPV evaluation point of view. The material is not homogeneous in reality. Such inhomogeneity comes in the effect of material inhomogeneity depending on the specimen location, heat treatment, and whole manufacturing process. The conventional master curve has a limitation to be applied to a large scatted data of fracture toughness such as the weld region

  6. Study on prediction model of irradiation embrittlement for reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Wang Rongshan; Xu Chaoliang; Huang Ping; Liu Xiangbing; Ren Ai; Chen Jun; Li Chengliang

    2014-01-01

    The study on prediction model of irradiation embrittlement for reactor pres- sure vessel (RPV) steel is an important method for long term operation. According to the deep analysis of the previous prediction models developed worldwide, the drawbacks of these models were given and a new irradiation embrittlement prediction model PMIE-2012 was developed. A corresponding reliability assessment was carried out by irradiation surveillance data. The assessment results show that the PMIE-2012 have a high reliability and accuracy on irradiation embrittlement prediction. (authors)

  7. Nano-structural changes in the RPV steels irradiated in MTR to high doses. 3D atom probe and positron annihilation study

    International Nuclear Information System (INIS)

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

    2005-01-01

    Reactor pressure vessel (RPV) steels of life-extended light water reactors are to be exposed to higher neutron fluence. The understanding of radiation embrittlement of RPV steels is very important in order to improve prediction of the embrittlement. The radiation embrittlement is mainly cased by copper-enriched cluster (CEC) and matrix damage (MD) due to irradiation. The state-or-the art technique such as three dimensional atom probe (3DAP) and positron annihilation (PA) has enabled to observe these microstructural features. The effect of highly dose irradiation on the formation of clusters in a low copper base metal and a high copper weld metal is investigated by means of the 3DAP and PA observations in this paper. The materials were irradiated to a neutron fluence of 10 20 n/cm 2 at 290 degC in a test reactor. The 3DAP observation shows that high dense CRCs in size of about 2 nm are formed in the high Cu weld metal. The CRCs consist of Si in addition to Fe, Cu, Mn, and Ni. Solute atom clusters below 2 nm are also observed in low Cu base metal, but the clusters include a large amount of Si and free from Cu. These clusters may be peculiar to highly irradiated materials because of no literature reporting such the clusters in the similar steels irradiated at the lower fluence. The data of the positron annihilation coincidence Doppler broadening measurement for both materials also shows the formation of clusters containing Cu, Ni, Mn, and Si. This means the clusters observed by 3DAP are uniformly distributed in the materials. Hardness tests and PA measurement combined with isochronal annealing show that defects, e.g. dislocation loop etc., having a positron lifetime of about 140 psec influence on mechanical properties of the steels. (author)

  8. Experimental device for investigating the crack growth behaviour of RPV steel under BWR conditions

    International Nuclear Information System (INIS)

    Anders, D.; Ahlf, J.

    1983-01-01

    An experimental device is developed to investigate the crack growth behaviour of RPV steel specimens under service conditions. It will be installed in the experimental power station VAK-Kahl (BWR, 16 MWe). The in pile part is composed of a stable frame with a hydraulically actuated load mechanism, the specimen chain and a measuring instrumentation. The specimen chain, fastened between load mechanism and a lower fixing point at the frame, is made up of five compact tensile specimens (CT40) and the associated connecting links. Specimen strain, crack opening and temperature are measured; for neutron dose monitoring activation wires are disposed. Out of pile, in the reactor hall, the hydraulic loading system is installed. The loading force is generated by a 100 kN-material testing machine; it moves a piston in the control cylinder, which is connected to the loading bellows of the in pile section. The measuring and control equipment and a desk computer serving for data preparation and reduction is placed in the reactor control room. (Auth.)

  9. FEM-calculation of different creep-tests with French and German RPV-steels

    International Nuclear Information System (INIS)

    Willschuetz, H.-G.; Altstadt, E.; Weiss, F.-P.; Sehgal, B.R.

    2003-01-01

    For calculations of Lower Head Failure experiments like FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. For an evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in 3 levels: starting with the simulation of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called 'tube-failure-experiments' are modeled: the RUPTHER-14 and the 'MPA-Meppen'- experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER experiments. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi55 RPV-steels, which are chemically nearly identical. If these 2 steels show a similar behavior, it should be allowed to transfer experimental and numerical data from one to the other. (author)

  10. Metallurgy of steels for PWR pressure vessels

    International Nuclear Information System (INIS)

    Kepka, M.; Mocek, J.; Barackova, L.

    1980-01-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure. (B.S.)

  11. Metallurgy of steels for PWR pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Kepka, M; Mocek, J; Barackova, L [Skoda, Plzen (Czechoslovakia)

    1980-09-01

    A survey and the chemical compositions are presented of reactor pressure vessel steels. The metallurgy is described of steel making for pressure vessels in Japan and the USSR. Both acidic and alkaline open-hearth steel is used for the manufacture of ingots. The leading world manufacturers of forging ingots for pressure vessels, however, exclusively use electric steel. Vacuum casting techniques are exclusively used. Experience is shown gained with the introduction of the manufacture of forging ingots for pressure vessels at SKODA, Plzen. The metallurgical procedure was tested utilizing alkaline open hearths, electric arc furnaces and facilities for vacuum casting of steel. Pure charge raw materials should be used for securing high steel purity. Prior to forging pressure vessel rings, not only should sufficiently big bottoms and heads be removed but also the ingot middle part should be scrapped showing higher contents of impurities and nonhomogeneous structure.

  12. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1996-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or open-quotes recovery,close quotes of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed

  13. Effects of thermal annealing and reirradiation on toughness of reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1997-01-01

    One of the options to mitigate the effects of irradiation on reactor pressure vessels (RPV) is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory (ORNL) to study the annealing response, or open-quotes recovery,close quotes of several irradiated RPV steels; it also includes recent results from both ORNL and the Russian Research Center-Kurchatov Institute (RRC-KI) on a cooperative program of irradiation, annealing and reirradiation of both U.S. and Russian RPV steels. The cooperative program was conducted under the auspices of Working Group 3, U.S./Russia Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS). The materials investigated are an RPV plate and various submerged-arc welds, with tensile, Charpy impact toughness, and fracture toughness results variously determined. Experimental results are compared with applicable prediction guidelines, while observed differences in annealing responses and reirradiation rates are discussed

  14. Miniature Precracked Charpy Specimens for Measuring the Master Curve Reference Temperature of RPV Steels at Impact Loading Rates

    Energy Technology Data Exchange (ETDEWEB)

    Lucon, E.; Scibetta, M.; Puzzolante, L.

    2008-10-15

    In the framework of the 2006 Convention, we investigated the applicability of fatigue precracked miniature Charpy specimens of KLST type (MPCC - B = 3 mm, W = 4 mm and L = 27 mm) for impact toughness measurements, using the well-characterized JRQ RPV steel. In the ductile to-brittle transition region, MPCC tests analyzed using the Master Curve approach and compared to data previously obtained from PCC specimens had shown a more ductile behavior and therefore un conservative results. In the investigation presented in this report, two additional RPV steels have been used to compare the performance of impact-tested MPCC and PCC specimens in the transition regime: the low-toughness JSPS steel and the high-toughness 20MnMoNi55 steel. The results obtained (excellent agreement for 20MnMoNi55 and considerable differences between T0 values for JSPS) are contradictory and do not presently allow qualifying the MPCC specimens as a reliable alternative to PCC samples for impact toughness measurements.

  15. Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

    International Nuclear Information System (INIS)

    Matlack, K. H.; Kim, J.-Y.; Wall, J. J.; Qu, J.; Jacobs, L. J.

    2014-01-01

    Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented

  16. Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

    Science.gov (United States)

    Matlack, K. H.; Kim, J.-Y.; Wall, J. J.; Qu, J.; Jacobs, L. J.

    2014-02-01

    Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented.

  17. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-01-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

  18. Critical cleavage fracture stress characterization of A508 nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Wu, Sujun; Jin, Huijin; Sun, Yanbin; Cao, Luowei

    2014-01-01

    The critical cleavage fracture stress of SA508 Gr.4N and SA508 Gr.3 low alloy reactor pressure vessel (RPV) steels was studied through the combination of experiments and finite element method (FEM) analysis. The results showed that the value of the local cleavage fracture stress, σ F , of SA508 Gr.4N steel was significantly higher than that of SA508 Gr.3 steel. Detailed microstructural analysis was carried out using FEGSEM which revealed much smaller grains, finer and more homogenous carbide particles formed in SA508 Gr.4N steel. Compared with the SA508 Gr.3 steel currently used in the nuclear industry, the SA508 Gr.4N steel possesses higher strength and notch toughness as well as improved cleavage fracture behavior, and is considered a better candidate RPV steel for the next generation nuclear reactors. - Highlights: • Critical cleavage fracture stress was calculated through experiments and FEM. • Effects of both grain and carbide particle sizes on σ F were discussed. • The SA508 Gr.4N steel is a better candidate for the next generation nuclear reactors

  19. Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Almjashev, V.I. [Alexandrov Scientific-Research Technology Institute (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [KTH, Stockholm (Sweden); Gusarov, V.V. [SPb State Technology University (SPbGTU), St. Petersburg (Russian Federation); Barrachin, M. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [EC-Joint Research Centre, Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Cadarache, St Paul lez Durance (France)

    2014-10-15

    Highlights: • The METCOR facility simulates vessel steel corrosion in contact with corium. • Steel corrosion rates in UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} coria accelerate above 1050 K. • However corrosion rates can also be limited by melt O{sub 2} supply. • The impact of this on in-vessel retention (IVR) strategy is discussed. - Abstract: During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 °C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR)

  20. Modeling of irradiation embrittlement and annealing/recovery in pressure vessel steels

    International Nuclear Information System (INIS)

    Lott, R.G.; Freyer, P.D.

    1996-01-01

    The results of reactor pressure vessel (RPV) annealing studies are interpreted in light of the current understanding of radiation embrittlement phenomena in RPV steels. An extensive RPV irradiation embrittlement and annealing database has been compiled and the data reveal that the majority of annealing studies completed to date have employed test reactor irradiated weldments. Although test reactor and power reactor irradiations result in similar embrittlement trends, subtle differences between these two damage states can become important in the interpretation of annealing results. Microstructural studies of irradiated steels suggest that there are several different irradiation-induced microstructural features that contribute to embrittlement. The amount of annealing recovery and the post-anneal re-embrittlement behavior of a steel are determined by the annealing response of these microstructural defects. The active embrittlement mechanisms are determined largely by the irradiation temperature and the material composition. Interpretation and thorough understanding of annealing results require a model that considers the underlying physical mechanisms of embrittlement. This paper presents a framework for the construction of a physically based mechanistic model of irradiation embrittlement and annealing behavior

  1. Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water

    International Nuclear Information System (INIS)

    Maruyama, Yu; Yamano, Norihiro; Moriyama, Kiyofumi; Park, Hyun Sun; Kudo, Tamotsu; Yang, Yanhua; Sugimoto, Jun

    1999-01-01

    In-vessel debris coolability experiments were performed in ALPHA program at JAERI. Molten aluminum oxide (Al 2 O 3 ) was poured into a pool of water in a lower head experimental vessel. Post-test observation and measurement using an ultrasonic technique indicated the formation of the interfacial gap between the solidified Al 2 O 3 and the vessel wall. Thermal responses of the vessel wall implied that the interfacial gap acted initially as a thermal resistance and water subsequently penetrated into the interfacial gap. The maximum heat flux at the inner surface of the vessel facing to the solidified Al 2 O 3 was roughly evaluated to be ranged from 320 kW/m 2 to 600 kW/m 2 . A post-test analysis was conducted with CAMP code. The influence of the interfacial gap on thermal behavior of Al 2 O 3 and the vessel wall was examined. (authors)

  2. 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)

  3. A physically-based correlation of irradiation-induced transition temperature shifts for RPV steels

    International Nuclear Information System (INIS)

    Eason, E.D.; Odette, G.R.; Nanstad, R.K.; Yamamoto, T.

    2013-01-01

    This paper presents a physically-based, empirically calibrated model for estimating irradiation-induced transition temperature shifts in reactor pressure vessel steels, based on a broader database and more complete understanding of embrittlement mechanisms than was available for earlier models. Brief descriptions of the underlying radiation damage mechanisms and the database are included, but the emphasis is on the model and the quality of its fit to U.S. power reactor surveillance data. The model is compared to a random sample of surveillance data that were set aside and not used in fitting and to selected independent data from test reactor irradiations, in both cases showing good ability to predict data that were not used for calibration. The model is a good fit to the surveillance data, with no significant residual error trends for variables included in the model or additional variables that could be included

  4. Effects of nickel on irradiation embrittlement of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    2005-06-01

    This TECDOC was developed under the IAEA Coordinated Research Project (CRP) entitled Effects of Nickel on Irradiation Embrittlement of Light Water Reactor Pressure Vessel (RPV) Steels. This CRP is the sixth in a series of CRPs to determine the influence of the mechanism and quantify the influence of nickel content on the deterioration of irradiation embrittlement of reactor pressure vessel steels of the Ni-Cr-Mo-V or Mn-Ni-Cr-Mo types. The scientific scope of the programme includes procurement of materials, determination of mechanical properties, irradiation and testing of specimens in power and/or test reactors, and microstructural characterization. Eleven institutes from eight different countries and the European Union participated in this CRP and six institutes conducted the irradiation experiments of the CRP materials. In addition to the irradiation and testing of those materials, irradiation experiments of various national steels were also conducted. Moreover, some institutes performed microstructural investigations of both the CRP materials and national steels. This TECDOC presents and discusses all the results obtained and the analyses performed under the CRP. The results analysed are clear in showing the significantly higher radiation sensitivity of high nickel weld metal (1.7 wt%) compared with the lower nickel base metal (1.2 wt%). These results are supported by other similar results in the literature for both WWER-1000 RPV materials, pressurized water reactor (PWR) type materials, and model alloys. Regardless of the increased sensitivity of WWER-1000 high nickel weld metal (1.7 wt%), the transition temperature shift for the WWER-1000 RPV design fluence is still below the curve predicted by the Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86). For higher fluence, no data were available and the results should not be extrapolated. Although manganese content was not incorporated directly in this CRP

  5. A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels

    Energy Technology Data Exchange (ETDEWEB)

    Eason, Ernest D. [Modeling and Computing Services, LLC; Odette, George Robert [UCSB; Nanstad, Randy K [ORNL; Yamamoto, Takuya [ORNL

    2007-11-01

    The reactor pressure vessels (RPVs) of commercial nuclear power plants are subject to embrittlement due to exposure to high-energy neutrons from the core, which causes changes in material toughness properties that increase with radiation exposure and are affected by many variables. Irradiation embrittlement of RPV beltline materials is currently evaluated using Regulatory Guide 1.99 Revision 2 (RG1.99/2), which presents methods for estimating the shift in Charpy transition temperature at 30 ft-lb (TTS) and the drop in Charpy upper shelf energy (ΔUSE). The purpose of the work reported here is to improve on the TTS correlation model in RG1.99/2 using the broader database now available and current understanding of embrittlement mechanisms. The USE database and models have not been updated since the publication of NUREG/CR-6551 and, therefore, are not discussed in this report. The revised embrittlement shift model is calibrated and validated on a substantially larger, better-balanced database compared to prior models, including over five times the amount of data used to develop RG1.99/2. It also contains about 27% more data than the most recent update to the surveillance shift database, in 2000. The key areas expanded in the current database relative to the database available in 2000 are low-flux, low-copper, and long-time, high-fluence exposures, all areas that were previously relatively sparse. All old and new surveillance data were reviewed for completeness, duplicates, and discrepancies in cooperation with the American Society for Testing and Materials (ASTM) Subcommittee E10.02 on Radiation Effects in Structural Materials. In the present modeling effort, a 10% random sample of data was reserved from the fitting process, and most aspects of the model were validated with that sample as well as other data not used in calibration. The model is a hybrid, incorporating both physically motivated features and empirical calibration to the U.S. power reactor surveillance

  6. 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

  7. Progress in Investigation of WWER-440 Reactor Pressure Vessel Steel by Gamma and Moessbauer Spectroscopy

    International Nuclear Information System (INIS)

    Hascik, J.; Slugen, V.; Lipka, J.; Hinca, R.; Toth, I.; Groene, R.; Uvacik, P.; Kupca, L.

    1998-01-01

    Gamma spectroscopic analyse and first experimental results of original irradiated reactor pressure vessel surveillance specimens are discussed in. In 1994, the new ''Extended Surveillance Specimen Program for nuclear Reactor Material Study'' was started in collaboration with the nuclear power plants (NPP) V-2 Bohunice (Slovakia). The first batch of MS samples (after 1 year, which is equivalent to 5 years of loading RPV-steel) was measured and interpreted using the new four components approach with the aim to observe microstructural changes due to thermal and neutron treatment resulting from operating conditions in NPP. The systematic changes in the relative areas of Moessbauer spectra components were observed. (author)

  8. Final report on the reactor pressure vessel pressurized-thermal-shock. International comparative assessment study (RPV PTS ICAS)

    International Nuclear Information System (INIS)

    Sievers, J.; Schulz, H.; Bass, R.; Pugh, C.

    1999-10-01

    A summary of the recently completed International Comparative Assessment Study of Pressurized-Thermal-Shock in Reactor Pressure Vessels (RPV PTS ICAS) is presented here to record the results in actual and comparative fashions. Within the DFM task, where account was taken of material properties and boundary conditions, reasonable agreement was obtained in linear-elastic and elastic-plastic analysis results. Linear elastic analyses and J-estimation schemes were shown to provide conservative estimates of peak crack driving force when compared with those obtained using complex three-dimensional (3D) finite element analyses. Predictions of RT NDT generally showed less scatter than that observed in crack driving force calculations due to the fracture toughness curve used for fracture assessment in the transition temperature region. Observed scatter in some analytical results could be traced mainly to a misinterpretation of the thermal expansion coefficient data given for the cladding and base metal. Also, differences in some results could be due to a quality assurance problem related to procedures for approximating the loading data given in the Problem Statement. For the PFM task, linear-elastic solutions were again shown to be conservative with respect to elastic-plastic solutions (by a factor of 2 to 4). Scatter in solutions obtained using the same computer code was generally attributable to differences in input parameters, e.g. standard deviations for the initial value of RT NDT , as well as for nickel and copper content. In the THM task, while there was a high degree of scatter during the early part of the transient, reasonable agreement in results was obtained during the latter part of the transient. Generally, the scatter was due to differences in analytical approaches used by participants, which included correlation-based engineering methods, system codes and three-dimensional computational fluids dynamics codes. Some of the models used to simulate condensation

  9. The effect of frequency and environment on the fatigue crack growth behaviour of SA508 Cl.III RPV steel

    International Nuclear Information System (INIS)

    Achilles, R.D.; Bulloch, J.H.

    1987-01-01

    This paper describes the effect of frequency and environment on the fatigue crack growth behaviour of SA508 Cl. III RPV steel. The study has shown that the effect of the Pressurised Water Reactor (PWR) environment is directly related to the frequency and the level of applied stress intensity of the test; these results further showed that the lower the frequency the greater the environmental effect, especially at low ΔK levels. No such frequency effect was observed in either the laboratory air or ultra-high purity argon environments. At a frequency of 0.1 Hz the PWR water test exhibited characteristic EAC growth, i.e. plateau growth behaviour. Fractographical examination of the fracture surface revealed that the fracture mode during plateau growth was intergranular failure. The experimental results are described and discussed in terms of the hydrogen assisted cracking mechanism. (author)

  10. Generation of a high temperature material data base and its application to creep tests with French or German RPV-steel. Technical report

    International Nuclear Information System (INIS)

    Willschuetz, H.G.; Altstadt, E.

    2002-08-01

    Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Numerous experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work /REM 1993/, /THF 1997/, /CHU 1999/. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties has been performed. For an evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in 3 levels: starting with the simulation of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called ''tube-failure-experiments'' are modeled: the RUPTHER-14 and the ''MPA-Meppen''-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. This report deals with the 1D- and 2D-simulations. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi55 RPV-steels, which are

  11. Effect of a Chloride Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 2)

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.-P.

    2002-11-01

    Within the CASTOC-project (5 t h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the second test of working package (WP) 3 with a NaCl transient, performed at Paul Scherrer Institut (PSI). In the first part of the experiment, an actively growing EAC crack with a crack growth rate (CGR) in the range of the 'low-sulphur SCC line' of the GE-model was generated by periodical partial unloading (PPU) in oxygenated high-temperature, high-purity water (T = 288 o C, DO = 8 ppm). Then a chloride transient of 49 ppb Cl - was applied for ∼40 h. After this transient, the load level was reduced and the loading conditions were changed to pure cyclic loading. Thereupon a second transient with a chloride concentration of 49 ppb was applied. In both RPV steels, the first chloride transient of 49 ppb Cl - resulted in an acceleration of the EAC crack growth by more than one order of magnitude and in fast, stationary SCC crack growth during the constant load phase of the PPU cycles at K I values 1/2 . 3 h after adding chloride to the high-purity water, the EAC CGR started to increase in the high-sulphur RPV steel during the constant load phase of a PPU cycle and after 20 h a stationary EAC CGR value in the range of the 'high-sulphur SCC curve' of the GE-model was reached. After 5 h in high-purity water, the crack growth began to slow down after a partial unloading cycle and 15 h later it reached again a stationary CGR value in the range of the 'low-sulphur SCC curve' of the GE-model. The second chloride transient did not result in an acceleration of the crack growth in both investigated specimens. This was explained by crack closure effects, which occurred in both specimens after the reduction of the load. The CGR

  12. 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

  13. Environmentally-Assisted Cracking of Low-Alloy Reactor Pressure Vessel Steels under Boiling Water Reactor Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Seifert, H.P.; Ritter, S

    2002-02-01

    The present report summarizes the experimental work performed by PSI on the environmentally-assisted cracking (EAC) of low-alloy steels (LAS) in the frame of the RIKORR-project during the period from January 2000 to August 2001. Within this project, the EAC crack growth behaviour of different low-alloy reactor pressure vessel (RPV) steels, weld filler and weld heat-affected zone materials is investigated under simulated transient and steady-state BWR/NWC power operation conditions. The EAC crack growth behaviour of different low-alloy RPV steels was characterized by slow rising load (SRL) / low-frequency corrosion fatigue (LFCF) and constant load tests with pre-cracked fracture mechanics specimens in oxygenated high-temperature water at temperatures of either 288, 250, 200 or 150 C. These tests revealed the following important interim results: Under low-flow and highly oxidizing (ECP >= 100 mV SHE) conditions, the ASME XI 'wet' reference fatigue crack growth curve could be significantly exceeded by cyclic fatigue loading at low frequencies (<0.001 Hz), at high and low load-ratios R, and by ripple loading near to DKth fatigue thresholds. The BWR VIP 60 SCC disposition lines may be significantly or slightly exceeded (even in steels with a low sulphur content) in the case of small load fluctuations at high load ratios (ripple loading) or at intermediate temperatures (200 -250 C) in RPV materials, which show a distinct susceptibility to dynamic strain ageing (DSA). (author)

  14. Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

    Science.gov (United States)

    Krasikov, E.; Nikolaenko, V.

    2017-01-01

    Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

  15. 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)

  16. The flow effect in the irradiation embrittlement in pressure vessel steels of nuclear power plants

    International Nuclear Information System (INIS)

    Kempf, Rodolfo A.; Cativa Tolosa, Sebastian; Fortis, Ana M.

    2009-01-01

    This paper deals with the advances in the study of the mechanical behavior of the Reactor Pressure Vessel steels under accelerate irradiations. The objective is to study the effect of lead factors on the interpretation of the mechanisms that induced the embrittlement of the RPV, like those of the reactors Atucha II and CAREM. It is described a device designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. It is presented also an automatic digital image processing technique for partitioning Charpy fracture surface into regions with a clear physical meaning and appropriate for the work in hot cells. The aim is to obtain the fracture behavior of irradiated specimens with different lead factors in the range of high fluencies and to know the dependence with the composition of the alloy and with the diffusion of other alloy elements. (author)

  17. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Reactor Pressure Vessels

    International Nuclear Information System (INIS)

    Potirniche, Gabriel; Barlow, Fred D.; Charit, Indrajit; Rink, Karl

    2013-01-01

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  18. Prediction and Monitoring Systems of Creep-Fracture Behavior of 9Cr-1Mo Steels for Teactor Pressure Vessels

    Energy Technology Data Exchange (ETDEWEB)

    Potirniche, Gabriel [Univ. of Idaho, Moscow, ID (United States); Barlow, Fred D. [Univ. of Idaho, Moscow, ID (United States); Charit, Indrajit [Univ. of Idaho, Moscow, ID (United States); Rink, Karl [Univ. of Idaho, Moscow, ID (United States)

    2013-11-26

    A recent workshop on next-generation nuclear plant (NGNP) topics underscored the need for research studies on the creep fracture behavior of two materials under consideration for reactor pressure vessel (RPV) applications: 9Cr-1Mo and SA-5XX steels. This research project will provide a fundamental understanding of creep fracture behavior of modified 9Cr-1Mo steel welds for through modeling and experimentation and will recommend a design for an RPV structural health monitoring system. Following are the specific objectives of this research project: Characterize metallurgical degradation in welded modified 9Cr-1Mo steel resulting from aging processes and creep service conditions; Perform creep tests and characterize the mechanisms of creep fracture process; Quantify how the microstructure degradation controls the creep strength of welded steel specimens; Perform finite element (FE) simulations using polycrystal plasticity to understand how grain texture affects the creep fracture properties of welds; Develop a microstructure-based creep fracture model to estimate RPVs service life; Manufacture small, prototypic, cylindrical pressure vessels, subject them to degradation by aging, and measure their leak rates; Simulate damage evolution in creep specimens by FE analyses; Develop a model that correlates gas leak rates from welded pressure vessels with the amount of microstructural damage; Perform large-scale FE simulations with a realistic microstructure to evaluate RPV performance at elevated temperatures and creep strength; Develop a fracture model for the structural integrity of RPVs subjected to creep loads; and Develop a plan for a non-destructive structural health monitoring technique and damage detection device for RPVs.

  19. Biaxial Loading Tests for steel containment vessel

    Energy Technology Data Exchange (ETDEWEB)

    Miyagawa, T. [Nuclear Power Engineering Corp., Tokyo (Japan); Wright, D.J.; Arai, S.

    1999-07-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  20. Biaxial Loading Tests for steel containment vessel

    International Nuclear Information System (INIS)

    Miyagawa, T.; Wright, D.J.; Arai, S.

    1999-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has conducted a 1/10 scale of the steel containment vessel (SCV) test for the understanding of ultimate structural behavior beyond the design pressure condition. Biaxial Loading Tests were supporting tests for the 1/10 scale SCV model to evaluate the method of estimating failure conditions of thin steel plates under biaxial loading conditions. The tentative material models of SGV480 and SPV490 were obtained. And the behavior of SGV480 and SPV490 thin steel plates under biaxial loading conditions could be well simulated by FE-Analyses with the tentative material models and Mises constitutive law. This paper describes the results and the evaluations of these tests. (author)

  1. RPV housed ATWS poison tank

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1992-01-01

    This patent describes a boiling water reactor (BWR) wherein housed within a reactor pressure vessel (RPV) is a nuclear core and an upper steam dome connected to a steam outlet in the RPV. The improvement comprises: a pressurized vessel disposed in the steam dome containing a neutron poison effective for inactivating the core and a first line for assaying the poison which first line runs to the outside of the RPV, the vessel being vented to the steam dome to pressurize the poison contained therein, the vessel being connected by a second line terminating beneath the core, the second line containing a valve which is actuable to release the poison through the line upon its actuation

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  3. The mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels: The case of Fe-Cu model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, A.V., E-mail: Alexey.V.Subbotin@gmail.com [Scientific and Production Complex Atomtechnoprom, Moscow 119180 (Russian Federation); Panyukov, S.V., E-mail: panyukov@lpi.ru [PN Lebedev Physics Institute, Russian Academy of Sciences, Moscow 117924 (Russian Federation)

    2016-08-15

    Mechanism of solute-enriched clusters formation in neutron-irradiated pressure vessel steels is proposed and developed in case of Fe-Cu model alloys. The suggested solute-drag mechanism is analogous to the well-known zone-refining process. We show that the obtained results are in good agreement with available experimental data on the parameters of clusters enriched with the alloying elements. Our model explains why the formation of solute-enriched clusters does not happen in austenitic stainless steels with fcc lattice structure. It also allows to quantify the method of evaluation of neutron irradiation dose for the process of RPV steels hardening.

  4. RPV steel embrittlement: Damage modeling and micro-mechanics in an engineering perspective

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A; Walle, E V; Chaouadi, R; Wannijn, J P; Werstrepen, A; Puzzolante, J L; VanRansbeeck, T H; VandeVelde, J [Sofia Univ. (Bulgaria)

    1994-12-31

    A new, consolidated strategy for improved Light Water Reactor pressure vessel surveillance is proposed. The methodology includes statistical fracture mechanics and damage modeling, while taking maximum advantage of the data generated by conventional surveillance practices. Available reconstitution and miniaturization allow to implement such strategy with minimal material inventory. The themes of the paper are: general philosophy of Belgian surveillance R D program; ductile-brittle transition temperature by use of instrumented C{sub v} load-time traces; towards an enhanced surveillance practice by combined use of instrumented C{sub v} load-time traces and uniaxial tensile tests; constraint, size and strain rate effects for C{sub v} notch impact test. 109 refs., 27 figs.

  5. Results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Luk, V.K.; Ludwigsen, J.S.; Hessheimer, M.F.; Komine, Kuniaki; Matsumoto, Tomoyuki; Costello, J.F.

    1998-05-01

    A series of static overpressurization tests of scale models of nuclear containment structures is being conducted by Sandia National Laboratories for the Nuclear Power Engineering Corporation of Japan and the US Nuclear Regulatory Commission. Two tests are being conducted: (1) a test of a model of a steel containment vessel (SCV) and (2) a test of a model of a prestressed concrete containment vessel (PCCV). This paper summarizes the conduct of the high pressure pneumatic test of the SCV model and the results of that test. Results of this test are summarized and are compared with pretest predictions performed by the sponsoring organizations and others who participated in a blind pretest prediction effort. Questions raised by this comparison are identified and plans for posttest analysis are discussed

  6. Pressure Vessel Steel Research: Belgian Activities

    International Nuclear Information System (INIS)

    Van Walle, E.; Fabry, A.; Ait Abderrahim, H.; Chaouadi, R.; D'hondt, P.; Puzzolante, J.L.; Van de Velde, J.; Van Ransbeeck, T.; Gerard, R.

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly

  7. Pressure Vessel Steel Research: Belgian Activities

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E; Fabry, A; Ait Abderrahim, H; Chaouadi, R; D` hondt, P; Puzzolante, J L; Van de Velde, J; Van Ransbeeck, T [Centre d` Etude de l` Energie Nucleaire, Mol (Belgium); Gerard, R [TRACTEBEL, Brussels (Belgium)

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly.

  8. Non-destructive evaluation of material degradation in RPV steel by magnetic methods

    International Nuclear Information System (INIS)

    Takahashi, S.; Kikuchi, H.; Kamada, Y.; Ara, K.; Zhang, L.; Liu, T.

    2004-01-01

    The minor hysteresis loops are measured with increasing magnetic field amplitude, H a , step by step and analyzed in connection with the lattice defects such as dislocations in deformed and neutron irradiated A533B steels. We have defined several new magnetic parameters in the minor loops: they are a pseudo coercive force H c *, a pseudo remanence B R *, a magnetic susceptibility at pseudo coercive force χ H *, pseudo hysteresis loss W f *, pseudo remanence work W r *. H c * is the magnetic field where the magnetization becomes zero in the minor loop. Six coefficients sensitive to lattice defects are obtained by the pseudo magnetic properties and they are independent of H a as well as the magnetic field. These coefficients are effective parameters for nondestructive evaluation of degradation before the initiation of cracking. The minor loops have several advantages for the nondestructive evaluation compared with the major loop. The coefficients have much information about lattice defects and the high accuracy. The measurement is available for low magnetic field of 20 Oe and the H a step is not necessarily fine for the detailed information because of the similarity. (orig.)

  9. On the transition of short cracks into long fatigue cracks in reactor pressure vessel steels

    Directory of Open Access Journals (Sweden)

    Singh Rajwinder

    2018-01-01

    Full Text Available Short fatigue cracks, having dimension less than 1 mm, propagate at much faster rates (da/dN even at lower stress intensity factor range (da/dN as compared to the threshold stress intensity factor range obtained from long fatigue crack growth studies. These short cracks originate at the sub-grain level and some of them ultimately transit into critical long cracks over time. Therefore, designing the components subjected to fatigue loading merely on the long crack growth data and neglecting the short crack growth behavior can overestimate the component’s life. This aspect of short fatigue cracks become even more critical for materials used for safety critical applications such as reactor pressure vessel (RPV steel in nuclear plants. In this work, the transition behaviour of short fatigue crack gowth into long fatigue crack is studied in SA508 Grade 3 Class I low alloy steel used in RPVs. In-situ characterization of initiation, propagation and transition of short fatigue cracks is performed using fatigue stage for Scanning Electron Microscope (SEM in addition to digital microscopes fitted over a servo-hydraulic fatigue machine and correlated with the microtructural information obtained using electron backscatter diffraction (EBSD. SA508 steel having an upper bainitic microstructure have several microstructural interfaces such as phase and grain boundaries that play a significant role in controlling the short fatigue crack propagation. Specially designed and prepared short fatigue specimens (eletro-polished with varying initial crack lengths of the order of tens of microns are used in this study. The transition of such short initial cracks into long cracks is then tracked to give detailed insight into the role of each phase and phase/grain boundary with an objective of establishing Kitagawa-Takahashi diagram for the given RPV steel. The behavior of the transited long cracks is then compared with the crack propagation behavior obtained using

  10. The IAEA data base ageing of reactor pressure vessel steels and welds

    International Nuclear Information System (INIS)

    Gillemot, F.; Ianko, L.; Davies, L.M.

    1995-01-01

    This paper describes one aspect of the International Atomic Energy Agency (IAEA) data base, that is to do with the ageing of reactor pressure vessel (RPV) steels and welds. It describes the background and the need for and the benefits deriving from such an international data base encompassing a greater number of sources than currently incorporated in existing international and national data bases. The paper describes the organization of this data base and the controls necessary for data acquisition and control. The current state of progress is described. Membership of and participation in this project is given and data access is also described. The technical features of the data base are described in terms of the structure of the data base and the hardware and software. New features are proposed such as the inclusion of measured curve data and metallographic data. Technical aspects of data evaluation are also included. (author). 1 ref., 6 figs

  11. Warm pre-stress experiments on highly irradiated reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Landron, C.; Ait-Bachir, M.; Moinereau, D.; Molinie, E.; Garbay, E.

    2015-01-01

    In the aim to justify in-service integrity of reactor pressure vessel beyond 40 years, experimental warm pre-stress (WPS) tests were performed on irradiated materials representative of RPV steels corresponding to 40 operating years. Different types of WPS loading path have been considered to cover typical postulated accidental transients. These results confirmed the beneficial effect of WPS on the cleavage fracture resistance of the irradiated materials. No fracture occurred during the cooling phase of the loading path and the fracture toughness values are higher than that measured with conventional isothermal tests. The analyses of the experiments, conducted using either simplified engineering models or more refined fracture models based on local approach to cleavage fracture, are in agreement with the experimental results. (authors)

  12. Influence of crack depth on the fracture toughness of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Theiss, T.J.; Bryson, J.W.

    1991-01-01

    The Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is investigating the influence of flaw depth on the fracture toughness of reactor pressure vessel (RPV) steel. Recently, it has been shown that, in notched beam testing, shallow cracks tend to exhibit an elevated toughness as a result of a loss of constraint at the crack tip. The loss of constraint takes place when interaction occurs between the elastic-plastic crack-tip stress field and the specimen surface nearest the crack tip. An increased shallow-crack fracture toughness is of interest to the nuclear industry because probabilistic fracture-mechanics evaluations show that shallow flaws play a dominant role in the probability of vessel failure during postulated pressurized-thermal-shock (PTS) events. Tests have been performed on beam specimens loaded in 3-point bending using unirradiated reactor pressure vessel material (A533 B). Testing has been conducted using specimens with a constant beam depth (W = 94 mm) and within the lower transition region of the toughness curve for A533 B. Test results indicate a significantly higher fracture toughness associated with the shallow flaw specimens compared to the fracture toughness determined using deep-crack (a/W = 0.5) specimens. Test data also show little influence of thickness on the fracture toughness for the current test temperature (-60 degree C). 21 refs., 5 figs., 3 tabs

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

    Science.gov (United States)

    Bergner, F.; Gillemot, F.; Hernández-Mayoral, M.; Serrano, M.; Török, G.; Ulbricht, A.; Altstadt, E.

    2015-06-01

    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.

  14. Biaxial loading effects on fracture toughness of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    McAfee, W.J.; Bass, B.R.; Bryson, J.W. Jr.; Pennell, W.E.

    1995-03-01

    The preliminary phases of a program to develop and evaluate fracture methodologies for assessing crack-tip constraint effects on fracture toughness of reactor pressure vessel (RPV) steels have been completed by the Heavy-Section Steel Technology (HSST) Program. Objectives were to investigate effect of biaxial loading on fracture toughness, quantify this effect through existing stress-based, dual-parameter, fracture-toughness correlations, or propose and verify alternate correlations. A cruciform beam specimen with 2-D, shallow, through-thickness flaw and a special loading fixture was designed and fabricated. Tests were performed using biaxial loading ratios of 0:1 (uniaxial), 0.6:1, and 1:1 (equi-biaxial). Critical fracture-toughness values were calculated for each test. Biaxial loading of 0.6:1 resulted in a reduction in the lower bound fracture toughness of ∼12% as compared to that from the uniaxial tests. The biaxial loading of 1:1 yielded two subsets of toughness values; one agreed well with the uniaxial data, while one was reduced by ∼43% when compared to the uniaxial data. Results were evaluated using J-Q theory and Dodds-Anderson (D-A) micromechanical scaling model. The D-A model predicted no biaxial effect, while the J-Q method gave inconclusive results. When applied to the 1:1 biaxial data, these constraint methodologies failed to predict the observed reduction in fracture toughness obtained in one experiment. A strain-based constraint methodology that considers the relationship between applied biaxial load, the plastic zone width in the crack plane, and fracture toughness was formulated and applied successfully to the data. Evaluation of this dual-parameter strain-based model led to the conclusion that it has the capability of representing fracture behavior of RPV steels in the transition region, including the effects of out-of-plane loading on fracture toughness. This report is designated as HSST Report No. 150

  15. Effect of a Sulphate Transient on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Test 1)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H. P

    2002-03-01

    Within the CASTOC-project (5th EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. Within WP 3 of this project, the Paul Scherrer Institut (PSI) investigates the effect of water chemistry transients on the EAC crack growth behaviour under periodical partial unloading (PPU) conditions. The present report is a summary of the first PSI test of WP 3 with a Na{sub 2}SO{sub 4} transient. In the first part of the report, the theoretical background on crack growth mechanisms, crack chemistry, mass transport and water chemistry transients as well as a brief literature survey on other water chemistry transient investigations is given. Furthermore, the experimental equipment and test procedure is presented, followed by a summary of the results of PSI test 1 of WP 3. Finally the results are discussed in detail and compared to literature data. In the first part of the experiment, an actively growing EAC crack was generated by PPU in oxygenated high-temperature, high-purity water (T = 288 {sup o}C, DO = 8 ppm, SO{sub 4}{sup 2-} < 0.6 ppb). Then a sulphate transient was applied. The duration ({approx} 300 h) and the amount of sulphate (SO{sub 4}{sup 2-} = 368 ppb) of the applied sulphate transient conservatively covered all sulphate transients, which might occur in BWR/normal water chemistry (NWC) practice. After the transient, outlet conductivity was lowered from ca. 1 {mu}S/cm to less than 0.15 {mu}S/cm within 2.6 h by a 'two-loop technique'. No accelerating effect of the sulphate transient on the EAC crack growth of both tested fracture mechanics specimens under highly oxidising BWR/NWC conditions was observed, making it impossible to deterrnine incubation or delay times. The EAC crack growth rates (CGR) before, during and after the

  16. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  17. A study on the irradiation embrittlement and recovery characteristics of light water reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Se Hwan; Hong, Jun Hwa; Lee, Bong Sang; Oh, Jong Myung; Song, Sook Hyang; Milan, Brumovsky [NRI Czech (Czech Republic)

    1999-03-01

    The neutron irradiation embrittlement phenomenon of light water RPV steels greatly affects the life span for safe operation of a reactor. Reliable evaluation and prediction of the embrittlement of RPV steels, especially of aged reactors, are of importance to the safe operation of a reactor. In addition, the thermal recovery of embrittled RPV has been recognized as an option for life extension. This study aimed to tracer/refine available technologies for embrittlement characterization and prediction, to prepare relevant materials for several domestic RPV steels of the embrittlement and recovery, and to find out possible remedy for steel property betterment. Small specimen test techniques, magnetic measurement techniques, and the Meechan and Brinkmann's recovery curve analysis method were examined/applied as the evaluation techniques. Results revealed a high irradiation sensitivity in YG 3 RPV steel. Further extended study may be urgently needed. Both the small specimen test technique for the direct determination of fracture toughness, and the magnetic measurement technique for embrittlement evaluation appeared to be continued for the technical improvement and data base preparation. Manufacturing process relevant to the heat treatment appeared to be improved in lowering the irradiation sensitivity of the steel. Further study is needed especially in applying the present techniques to the new structural materials under new irradiation environment of advanced reactors. (author)

  18. 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)

  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. Specific Features of Structural-Phase State and Properties of Reactor Pressure Vessel Steel at Elevated Irradiation Temperature

    Directory of Open Access Journals (Sweden)

    E. A. Kuleshova

    2017-01-01

    Full Text Available This paper considers influence of elevated irradiation temperature on structure and properties of 15Kh2NMFAA reactor pressure vessel (RPV steel. The steel is investigated after accelerated irradiation at 300°C (operating temperature of VVER-1000-type RPV and 400°C supposed to be the operating temperature of advanced RPVs. Irradiation at 300°C leads to formation of radiation-induced precipitates and radiation defects-dislocation loops, while no carbide phase transformation is observed. Irradiation at a higher temperature (400°C neither causes formation of radiation-induced precipitates nor provides formation of dislocation loops, but it does increase the number density of the main initial hardening phase—of the carbonitrides. Increase of phosphorus concentration in grain boundaries is more pronounced for irradiation at 400°C as compared to irradiation at 300°C due to influence of thermally enhanced diffusion at a higher temperature. The structural-phase changes determine the changes of mechanical properties: at both irradiation temperatures irradiation embrittlement is mainly due to the hardening mechanism with some contribution of the nonhardening one for irradiation at 400°C. Lack of formation of radiation-induced precipitates at T = 400°C provides a small ΔTK shift (17°C. The obtained results demonstrate that the investigated 15Kh2NMFAA steel may be a promising material for advanced reactors with an elevated operating temperature.

  1. Topic 1. Steels for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Brynda, J.; Kepka, M.; Barackova, L.; Vacek, M.; Havel, S.; Cukr, B.; Protiva, K.; Petrman, I.; Tvrdy, M.; Hyspecka, L.; Mazanec, K.; Kupca, L.; Brezina, M.

    1980-01-01

    Part 1 of the Proceedings consists of papers on the criteria for the selection and comparison of the properties of steel for pressure vessels and on the metallurgy of the said steels, the selection of suitable material for internal tubing systems, the manufacture of high-alloy steels for WWER components, the mechanical and metallurgical properties of steel 22K for WWER 440 pressure components, and of steel 10MnNi2Mo for the WWER primary coolant circuit, and the metallographic assessment of steel 0Kh18N10T. (J.P.)

  2. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheres during an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities and oxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium-specimen ingot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction

  3. Defects investigation in neutron irradiated reactor steels by positron annihilation

    International Nuclear Information System (INIS)

    Slugen, V.

    2003-01-01

    Positron annihilation spectroscopy (PAS) based on positron lifetime measurements using the Pulsed Low Energy Positron System (PLEPS) was applied to the investigation of defects of irradiated and thermally treated reactor pressure vessel (RPV) steels. PLEPS results showed that the changes in microstructure of the RPV-steel properties caused by neutron irradiation and post-irradiation heat treatment can be well detected. From the lifetime measurements in the near-surface region (20-550 nm) the defect density in Russian types of RPV-steels was calculated using the diffusion trapping model. The post-irradiation heat treatment studies performed on non-irradiated specimens are also presented. (author)

  4. Experimental studies of oxidic molten corium-vessel steel interaction

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V.

    2001-01-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere

  5. Experimental studies of oxidic molten corium-vessel steel interaction

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. E-mail: niti-npc@sbor.net; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V

    2001-12-01

    The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

  6. Low temperature radiation embrittlement for reactor vessel steels

    International Nuclear Information System (INIS)

    Ginding, I.A.; Chirkina, L.A.

    1978-01-01

    General conceptions of cold brittleness of bcc metals are in a review. Considered are experimental data and theoretical representations about the effect of irradiation conditions, chemical composition, phase and structural constitutions, grain size, mechanical and thermomechanical treatments on low-temperature irradiation embrittlement of reactor vessel steels. Presented are the methods for increasing radiation stability of metals (carbon and Cr-Mo steels) used in manufacturing reactor vessels

  7. Formation mechanism of solute clusters under neutron irradiation in ferritic model alloys and in a reactor pressure vessel steel: clusters of defects

    International Nuclear Information System (INIS)

    Meslin-Chiffon, E.

    2007-11-01

    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)

  8. Application of high strength steel to nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Susukida, H.; Sato, M.; Takano, G.; Uebayashi, T.; Yoshida, K.

    1976-01-01

    Nuclear reactor containment vessels are becoming larger in size with the increase in the power generating capacity of nuclear power plants. For example, a containment vessel for a PWR power plant with an output of 1,000 MWe becomes an extremely large one if it is made of the conventional JIS SGV 49 (ASTM A 516 Gr. 70) steel plates less than 38 mm in thickness. In order to design the steel containment vessel within the conventional dimensional range, therefore, it is necessary to use a high strength steel having a higher tensile strength than SGV 49 steel, good weldability and a higher fracture toughness and moreover, possessing satisfactory properties without undergoing post-weld heat treatment. The authors conducted a series of verification tests on high strength steel developed by modifying the ASTM A 543 Grade B Class 1 steel with a view to adopting it as a material for the nuclear reactor containment vessels. As the result of evaluation of the test results from various angles, we confirmed that the high strength steel is quite suitable for the manufacture of nuclear reactor containment vessels. (auth.)

  9. Status report on the use of the CRB for the measurement of fracture toughness of RPV steels

    International Nuclear Information System (INIS)

    Scibetta, M.; Chaouadi, R.; Van Walle, E.

    1998-02-01

    A large number of fracture toughness tests were performed in order to assess the use of the circumferentially-Cracked Round Bar (CRB) as a potential method for the measurement of fracture toughness of Reactor Pressure Vessel steels. Test conditions were selected to: (1) characterise fracture toughness in the transition region; (2) study the size effect and loss of constraint; (3) establish the limit of validity of this geometry; (4) investigate the ductile fracture at the upper shelf. In the transition region, the fracture toughness obtained from the CRB over-estimates the actual value as long as the loss of constraint and size effect were not taken into account. In addition, the B1/4 size correction is verified and gives a very good description of the size effect. The application of these corrections allows a good prediction of the normalised fracture toughness up to high levels of fracture toughness.In the upper shelf region, promising results were obtained with this geometry to characterise the ductile crack initiation and propagation

  10. Evaluation of neutron irradiation embrittlement in the Korean reactor pressure vessel steels (Final Report)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, J. H.; Lee, B. S.; Chi, S. H.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Kwon, S. C.; Park, D. G.; Kang, Y. H.; Choo, K. N.; Oh, J. M.; Park, S. J.; Kim, B. K.; Shin, Y. T.; Cho, M. S.; Sohn, J. M.; Kim, D. S.; Choo, Y. S.; Ahn, S. B.; Oh, W. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-05-01

    Reactor pressure vessel materials, which were produced by a domestic company, Doosan Heavy Industries and construction Co., Ltd., have been evaluated using the neutron irradiation facility HANARO. For this evaluation, instrumented capsules were used for neutron irradiation of various kinds of specimens made of different heats of steels, which are VCD(Y4), VCD+Al(U4), Si+Al(Y5), U4 weld metal, and U4 HAZ, respectively. The fast neutron fluence levels ranged 1 to 5 (x10{sup 19} n/cm{sup 2}, E>1MeV) depending on the specimens and the irradiation temperature was controlled within 290{+-}10 deg C. The test results showed that, in the ranking of the material properties of the base metals, both before and after neutron irradiation, Y5 is the best, U4 the next and Y4 the last. Y4 showed a substantial change by neutron irradiation as well as the properties was worse than others in the unirradiated state. However, Y5, which showed the best properties in unirradiated state, was also the best in the resistance for irradiation embrittlement and one can hardly detect the property change after irradiation. The weldment showed a reasonably good resistance to irradiation embrittlement while the unirradiated properties were worse than base metals. The RPV steels are all expected to meet the screening criteria of the USNRC codes and regulations during the end of plant life. 39 refs., 42 figs., 27 tabs. (Author)

  11. 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.

  12. Problems in development of pressure vessel steels

    International Nuclear Information System (INIS)

    McMahon, C.Y.

    1980-01-01

    The tendency of steels to intercrystalline fracture at low stresses is the main factor, limiting fracture resistance of steels in agressive media at conventional and elevated temperatures. The reasons for the phenomenon are analyzed. In particular, the role of grain boundary segregations of non-metallic impurities is pointed out. The ways of the problem solving both at the expense of corresponding microstructure control and by means of selection of the steel chemical composition are considered

  13. Small specimen measurements of dynamic fracture toughness of heavy section steels for nuclear pressure vessel

    International Nuclear Information System (INIS)

    Tanaka, Y.; Iwadate, T.; Suzuki, K.

    1987-01-01

    This study presents the dynamic fracture toughness properties (KId) of 12 heats of RPV steels measured using small specimens and analysed based on the current research. The correlation between the KId test and other engineering small specimen tests such as Charpy test and drop weight test are also discussed and a method to predict the KId value is presented. (orig./HP)

  14. Applicability of the fracture toughness master curve to irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Sokolov, M.A.; McCabe, D.E.; Alexander, D.J.; Nanstad, R.K.

    1997-01-01

    The current methodology for determination of fracture toughness of irradiated reactor pressure vessel (RPV) steels is based on the upward temperature shift of the American Society of Mechanical Engineers (ASME) K Ic curve from either measurement of Charpy impact surveillance specimens or predictive calculations based on a database of Charpy impact tests from RPV surveillance programs. Currently, the provisions for determination of the upward temperature shift of the curve due to irradiation are based on the Charpy V-notch (CVN) 41-J shift, and the shape of the fracture toughness curve is assumed to not change as a consequence or irradiation. The ASME curve is a function of test temperature (T) normalized to a reference nit-ductility temperature, RT NDT , namely, T-RT NDT . That curve was constructed as the lower boundary to the available K Ic database and, therefore, does not consider probability matters. Moreover, to achieve valid fracture toughness data in the temperature range where the rate of fracture toughness increase with temperature is rapidly increasing, very large test specimens were needed to maintain plain-strain, linear-elastic conditions. Such large specimens are impractical for fracture toughness testing of each RPV steel, but the evolution of elastic-plastic fracture mechanics has led to the use of relatively small test specimens to achieve acceptable cleavage fracture toughness measurements, K Jc , in the transition temperature range. Accompanying this evolution is the employment of the Weibull distribution function to model the scatter of fracture toughness values in the transition range. Thus, a probabilistic-based bound for a given data population can be made. Further, it has been demonstrated by Wallin that the probabilistic-based estimates of median fracture toughness of ferritic steels tend to form transition curves of the same shape, the so-called ''master curve'', normalized to one common specimen size, namely the 1T [i.e., 1.0-in

  15. Irradiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Bros, J.

    2000-01-01

    From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

  16. Time-dependent temper embrittlement of reactor pressure vessel steel: Correlation between microstructural evolution and mechanical properties during tempering at 650 °C

    Energy Technology Data Exchange (ETDEWEB)

    Li, Chuanwei; Han, Lizhan; Yan, Guanghua; Liu, Qingdong; Luo, Xiaomeng; Gu, Jianfeng, E-mail: gujf@sjtu.edu.cn

    2016-11-15

    The microstructural evolution of reactor pressure vessel (RPV) steel and its effect on the mechanical properties during tempering at 650 °C were studied to reveal the time-dependent toughness and temper embrittlement. The results show that the toughening of the material should be attributed to the decomposition of the martensite/austenite constituents and uniform distribution of carbides. When the tempering duration was 5 h, the strength of the investigated steel decreased to strike a balance with the material impact toughness that reached a plateau. As the tempering duration was further increased, the material strength was slightly reduced but the material impact toughness deteriorated drastically. This time-dependent temper embrittlement is different from traditional temper embrittlement, and it can be partly attributed to the softening of the matrix and the broadening of the ferrite laths. Moreover, the dimensions and distribution of the grain carbides are the most important factors of the impact toughness. - Highlights: • The fracture mechanism of reactor pressure vessel (RPV) steels under impact load was investigated. • The Charpy V-notch impact test and the hinge model were employed for the study. • Grain boundary carbides play a key role in the impact toughness and fracture toughness. • The dependence of the deterioration of impact toughness on tempering time was analyzed for the first time.

  17. Corrosion of vessel steel during its interaction with molten corium

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

    2006-01-01

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments

  18. Corrosion of vessel steel during its interaction with molten corium

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation)]. E-mail: bechta@sbor.spb.su; Khabensky, V.B. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Vitol, S.A. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Krushinov, E.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Granovsky, V.S. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Lopukh, D.B. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Gusarov, V.V. [Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), Odoevsky str., b. 24/2, 199155 St. Petersburg (Russian Federation); Martinov, A.P. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Martinov, V.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Fieg, G. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Tromm, W. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Bottomley, D. [Europaeische Kommission, General Direktion GFS, Institut fuer Transurane (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tuomisto, H. [Fortum Engineering Ltd. 00048 FORTUM, Rajatorpantie 8, Vantaa (Finland)

    2006-07-15

    An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments.

  19. Radiation embrittlement of Spanish nuclear reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Bros, J.; Ballesteros, A.; Lopez, A.

    1993-01-01

    Commercial pressurized water reactor (PWR) and boiling water reactor (BWR) nuclear power plants contain a series of pressure vessel steel surveillance capsules as the principal means of monitoring radiation effects on the pressure vessel. Changes in fracture toughness are more severe in surveillance capsules than in reactor vessel materials because of their proximity of the reactor core. Therefore, it is possible to predict changes in fracture toughness of the reactor vessel materials. This paper describes the status of the reactor vessel surveillance program relating to Spanish nuclear power plants. To date, twelve capsules have been removed and analyzed from seven of the nine Spanish reactors in operation. The results obtained from the analysis of these capsules are compared with the predictions of the Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Rev. 2, by means of measured and expected increase of the nil-ductility transition reference temperature (RT NDT ). The comparison is made considering the different variables normally included in the studies of radiation response of reactor pressure vessel materials, such as copper content of steel, level of neutron fluence above 1 MeV, base metal or weld metal, and so forth. The surveillance data have been used for determining the adjusted reference temperatures and upper shelf energies at any time. The results have shown that the seven pressure vessels are in excellent condition to continue operating with safety against brittle fracture beyond the design life, without the need to recuperate the degraded properties of the materials by annealing of the vessel

  20. 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

  1. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-01-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program

  2. Testing of a steel containment vessel model

    International Nuclear Information System (INIS)

    Luk, V.K.; Hessheimer, M.F.; Matsumoto, T.; Komine, K.; Costello, J.F.

    1997-01-01

    A mixed-scale containment vessel model, with 1:10 in containment geometry and 1:4 in shell thickness, was fabricated to represent an improved, boiling water reactor (BWR) Mark II containment vessel. A contact structure, installed over the model and separated at a nominally uniform distance from it, provided a simplified representation of a reactor shield building in the actual plant. This paper describes the pretest preparations and the conduct of the high pressure test of the model performed on December 11-12, 1996. 4 refs., 2 figs

  3. On the composition and structure of nanoprecipitates in irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Odette, G.R.; Liu, C.L.; Wirth, B.D.

    1997-01-01

    Nanoscale Cu rich precipitates (CRPs) are widely believed to be the dominant hardening feature resulting in severe embrittlement in irradiated reactor pressure vessel (RPV) steels. However, this view has recently been challenged by interpretations of atom probe field ion microscopy (APFIM) measurements that describe the dominant nanofeatures as dilute solute atmospheres (DSAs). The practical impact of these differing views is very significant. This work compares and contrasts the CRP versus DSA descriptions to a wide variety of pertinent data. Mechanical property trends as well as small angle neutron scattering (SANS) and field emission scanning transmission electron microscopy (FEGSTEM) measurements support the presence of CRPs. CRPs are also consistent with the fundamental thermodynamic and kinetic laws. However, standard theory cannot provide the atomic level resolution needed to fully understand the nanofeatures. Therefore, a new Lattice Monte Carlo (LMC) atomistic method is used to simulate the complex chemical structures of the CRPs. The LMC method unifies the SANS/FEGSTEM and APFIM data within a well founded physical framework

  4. Heavy-Section Steel Irradiation Program

    Energy Technology Data Exchange (ETDEWEB)

    Rosseel, T.M.

    2000-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established.

  5. Heavy-Section Steel Irradiation Program

    International Nuclear Information System (INIS)

    Rosseel, T.M.

    2000-01-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established

  6. 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

  7. Fracture Toughness Evaluation of Kori-1 RPV Beltline Weld for a Long-Term Operation

    International Nuclear Information System (INIS)

    Lee, Bong-Sang; Kim, Min-Chul; Ahn, Sang-Bok; Kim, Byung-Chul; Hong, Jun-Hwa

    2007-01-01

    Irradiation embrittlement of RPV (reactor pressure vessel) material is the most important aging issue for a long-term operation of nuclear power plants. KORI unit 1, which is the first PWR in Korea, is approaching its initial licensing life of 30 years. In order to operate the reactor for another 10 years and more, it should be demonstrated that the irradiation embrittlement of the reactor will be adequately managed by ensuring that the fracture toughness properties have a certain level of the safety margin. The current regulation requires Charpy V-notch impact data through conventional surveillance tests. It is based on the assumption that Charpy impact test results are well correlated with the fracture toughness properties of many engineering steels. However, Charpy V-notch impact data may not be adequate to estimate the fracture toughness of certain materials, such as Linde 80 welds. During the last decade, a tremendous number of fracture toughness data on many RPV steels have been produced in accordance with the new standard test method, the so-called master curve method. ASTM E1921 represents a revolutionary advance in characterizing fracture toughness of RPV steels, since it permits establishing the ductile to brittle transition portion of the fracture toughness curve with direct measurements on a relatively small number of relatively small specimens, such as pre-cracked Charpy specimens. Actual fracture toughness data from many different RPV steels revealed that the Charpy test estimations are generally conservative with the exception of a few cases. Recent regulation codes in USA permit the master curve fracture toughness methodology in evaluating an irradiation embrittlement of commercial nuclear reactor vessels

  8. Prediction of the brittle fracture toughness value of a RPV steel from the analysis of a limited set of Charpy results

    International Nuclear Information System (INIS)

    Forget, P.; Marini, B.; Verdiere, N.

    2001-01-01

    Our objective is to establish a method to be able to determine fracture toughness of a reactor pressure vessel (RPV) by using the small number of Charpy specimens used in the reactor surveillance program. Previous studies have shown that it is possible to determine fracture toughness from Charpy tests. Another point is to determine if statistical effects are compatible with a restricted number of specimens, this paper deals with this point and presents a methodology that is applicable to the case of irradiated materials from the surveillance program. Several conclusions can be drawn from this study: -) When determining failure parameters, we gain most accuracy by increasing the number of samples from 3 to about 6; -) it is possible to evaluate brittle fracture toughness using local approach, either by using Beremin or Renevey model; -) The effect of using a small number of Charpy specimens to determine fracture toughness in brittle fracture is evaluated. The error in the evaluation of fracture toughness is much smaller than the experimental dispersion itself. (A.C.)

  9. Corrosion fatigue of pressure vessel steels in PWR environments--influence of steel sulfur content

    International Nuclear Information System (INIS)

    Scott, P.M.; Druce, S.G.; Truswell, A.E.

    1984-01-01

    Large effects of simulated light water reactor environments at 288 C on fatigue crack growth in low alloy pressure vessel steels are observed only when specific mechanical, metallurgical, and electrochemical conditions are satisfied simultaneously. In this paper, the relative importance of three key variables--steel impurity content, water chemistry, and flow rate--and their interaction with loading rate or strain rate are examined. In particular, the results of a systematic examination of the influence of a steel's sulfur content are described

  10. 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.

  11. Fracture toughness of irradiated and recovered vessel steels

    International Nuclear Information System (INIS)

    Perosanz, F.; Lapena, J.

    1998-01-01

    This paper presents the fracture toughness measurements carried out on three vessel steels in an irradiated condition and after a post-irradiation recovery treatment. A statistical approach and the fracture parameters corresponding to two theoretical models of the fracture tests are used for evaluating toughness. Test results show that the neutron fluence gradually transforms the fracture behaviour of the vessel steels from ductile to brittle and seriously reduces their fracture toughness. The effectiveness of the recovery treatment, as evaluated from the toughness measurements, is confirmed, although the efficiency is not the same for the steels and depends on the evaluation parameter except in the case of almost complete recovery. The recovery effect increases with the received neutron fluence if the toughness values after treatment are compared with those in the irradiated condition rather than those in the as received condition. (orig.)

  12. Microstructural evolution in reactor pressure vessel steel under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Ohno, Katsumi; Fukuya, Koji [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Understanding microstructural changes in reactor pressure vessel steels is important in order to evaluate radiation-induced embrittlement, one of the major aging phenomena affecting the extension of plant life. In this study, actual surveillance test specimens and samples of rector vessel low-alloy steel (A533B steel) irradiated in a research reactor were examined using state-of-the-art techniques to clarify the neutron flux effect on the microstructural changes. These techniques included small angle neutron scattering and atom probes. Microstructural changes which are considered to be the main factors affecting embrittlement, including the production of copper-rich precipitates and the segregation of impurity elements, were confirmed by the results of the study. In addition, the mechanical properties were predicted based on the obtained quantitative data such as the diameters of precipitates. Consequently, the hardening due to irradiation was almost simulated. (author)

  13. 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)

  14. 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)

  15. 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.)

  16. The annal of british RPV steel plates for first nuclear power station in Japan (1). Unforseen accidents araised before nuclear power plant open

    International Nuclear Information System (INIS)

    Miyoshi, Shigeru

    2011-01-01

    This article described the author's experiences of reactor vessel steel plates for the first nuclear power station, Tokai-mura reactor. The station is an advanced Calder Hall type. The electrical output is 166 MWe. The reactor vessel was spherical with internal diameter of 189 cm and wall thickness of 83 mm. Material was a fine-grain, aluminum-killed steel. Each part of pressure vessel, bottom cap, belt 1, 2, 3, 4 and top cap, were prefabricated with welding of plates, then lifted into the reactor building and assembled with welding. Steel plates were imported from UK, press formed to spherical segments in Japan and transferred to the site. Ultrasonic testing, magnetic particle testing of groove face (crack detection), sizing of groove and sulfur print tests were performed as an on-site acceptance testing. Inclusions and lamination openings were observed at groove faces due to gas flame cutting. White spot was observed at rupture face of tensile test specimen. At the liquid penetration testing after back gauging of extra seam, a crack-like indication with length of less than 3 mm was observed. Reexamination of groove face by magnetic particles testing showed indications of inclusion cloud or alumina cloud. These would be cracks caused by hydrogen embrittlement. (T. Tanaka)

  17. Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

    International Nuclear Information System (INIS)

    Whitman, G.D.

    1975-01-01

    The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references

  18. Study on Material Selection of Reactor Pressure Vessel of SCWR

    Science.gov (United States)

    Ma, Shuli; Luo, Ying; Yin, Qinwei; Li, Changxiang; Xie, Guofu

    This paper first analyzes the feasibility of SA-508 Grade 3 Class 1 Steel as an alternative material for Supercritical Water-Cooled Reactor (SCWR) Reactor Pressure Vessel (RPV). This kind of steel is limited to be applied in SCWR RPV due to its quenching property, though large forging could be accomplished by domestic manufacturers in forging aspect. Therefore, steels with higher strength and better quenching property are needed for SWCR RPV. The chemical component of SA-508 Gr.3 Cl.2 steel is similar to that of SA-508 Gr.3 Cl.1 steel, and more appropriate matching of strength and toughness could be achieved by the adjusting the elements contents, as well as proper control of tempering temperature and time. In light of the fact that Cl.2 steel has been successfully applied to steam generator, it could be an alternative material for SWCR RPV. SA-508 Gr.4N steel with high strength and good toughness is another alternative material for SCWR RPV. But large amount of research work before application is still needed for the lack of data on welding and irradiation etc.

  19. Magnetic and electrical properties of ITER vacuum vessel steels

    International Nuclear Information System (INIS)

    Mergia, K.; Apostolopoulos, G.; Gjoka, M.; Niarchos, D.

    2007-01-01

    Full text of publication follows: Ferritic steel AISI 430 is a candidate material for the lTER vacuum vessel which will be used to limit the ripple in the toroidal magnetic field. The magnetic and electrical properties and their temperature dependence in the temperature range 300 - 900 K of AISI 430 ferritic stainless steels are presented. The temperature variation of the coercive field, remanence and saturation magnetization as well as electrical resistivity and the effect of annealing on these properties is discussed. (authors)

  20. Neutron irradiation effects in pressure vessel steels and weldments

    Energy Technology Data Exchange (ETDEWEB)

    Ianko, L [International Atomic Energy Agency, Vienna (Austria). Div. of Nuclear Power; Davies, L M

    1994-12-31

    This paper deals with the effects of neutron irradiation on the steel and welds used for the pressure vessels which house the reactor cores in light water reactors: irradiation effects on mechanical properties and the shift in ductile-brittle transition temperature, importance of the knowledge of the neutron fluence and of the monitoring and surveillance programmes; empirical and mechanistic modelling of irradiation effects and the necessity of data extension to new operational limits; consequences on the manufacturing and structural design of materials and structures; mitigation of irradiation effects by annealing; international activities and programmes in the field of neutron irradiation effects on PV steels and welds. 37 refs., 22 figs.

  1. Fracture mechanics analysis and evaluation for the RPV of the Chinese Qinshan 300 MW NPP and PTS

    International Nuclear Information System (INIS)

    He Yinbiao; Isozaki, Toshikuni

    2000-03-01

    One of the most severe accident conditions of a reactor pressure vessel (RPV) in service is the loss of coolant accident (LOCA). Cold safety injection water is pumped into the downcomer of the RPV through inlet nozzles, while the internal pressure may remain at high level. Such an accident is called pressurized thermal shock (PTS) transient according to 10 CFR 50.61 definition. This paper illustrates the fracture mechanics analysis for the existing RPV of the Chinese Qinshan 300 MW nuclear power plant (NPP) under the postulated PTS transients that include SB-LOCA, LB-LOCA of Qinshan NPP and Rancho Seco transients. 3-D models with the flaw depth range a/w=0.05∼0.9 (a: flaw depth; w: wall thickness) were used to probe what kind of flaw and what kind of transient are most dangerous for the RPV in the end of life (EOF). Both the elastic and elastic-plastic material models were used in the stress analysis and fracture mechanics analysis. The different types of flaw and the influence of the stainless steel cladding on the fracture analysis were investigated for different PTS transients. comparing with the material initiation crack toughness K IC , the fracture evaluation for the RPV in question under PTS transients are performed in this paper. (author)

  2. Application of the master curve approach to fracture mechanics characterisation of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Viehrig, Hans-Werner; Zurbuchen, Conrad; Kalkhof, Dietmar

    2010-06-01

    The paper presents results of a research project founded by the Swiss Federal Nuclear Inspectorate concerning the application of the Master Curve approach in nuclear reactor pressure vessels integrity assessment. The main focus is put on the applicability of pre-cracked 0.4T-SE(B) specimens with short cracks, the verification of transferability of MC reference temperatures T 0 from 0.4T thick specimens to larger specimens, ascertaining the influence of the specimen type and the test temperature on T 0 , investigation of the applicability of specimens with electroerosive notches for the fracture toughness testing, and the quantification of the loading rate and specimen type on T 0 . The test material is a forged ring of steel 22 NiMoCr 3-7 of the uncommissioned German pressurized water reactor Biblis C. SE(B) specimens with different overall sizes (specimen thickness B=0.4T, 0.8T, 1.6T, 3T, fatigue pre-cracked to a/W=0.5 and 20% side-grooved) have comparable T 0 . T 0 varies within the 1σ scatter band. The testing of C(T) specimens results in higher T 0 compared to SE(B) specimens. It can be stated that except for the lowest test temperature allowed by ASTM E1921-09a, the T 0 values evaluated with specimens tested at different test temperatures are consistent. The testing in the temperature range of T 0 ± 20 K is recommended because it gave the highest accuracy. Specimens with a/W=0.3 and a/W=0.5 crack length ratios yield comparable T 0 . The T 0 of EDM notched specimens lie 41 K up to 54 K below the T 0 of fatigue pre-cracked specimens. A significant influence of the loading rate on the MC T 0 was observed. The HSK AN 425 test procedure is a suitable method to evaluate dynamic MC tests. The reference temperature T 0 is eligible to define a reference temperature RT To for the ASME-KIC reference curve as recommended in the ASME Code Case N-629. An additional margin has to be defined for the specific type of transient to be considered in the RPV integrity assessment

  3. Radiation embrittlement of WWER 440 pressure vessel steel and of some improved steels by western producers

    International Nuclear Information System (INIS)

    Koutsky, J.; Vacek, M.; Stoces, B.; Pav, T.; Otruba, J.; Novosad, P.; Brumovsky, M.

    1982-01-01

    The resistance was studied of Cr-Mo-V type steel 15Kh2MFA to radiation embrittlement at an irradiation temperature of around 288 degC. Studied was the steel used for the manufacture of the pressure vessel of the Paks nuclear reactor in Hungary. The obtained results of radiation embrittlement and hardening of steel 15Kh2MFA were compared with similar values of Mn-Ni-Mo type steels A 533-B and A 508 manufactured by leading western manufacturers within the international research programme coordinated by the IAEA. It was found that the resistance of steel 15Kh2MFA to radiation embrittlement is comparable with steels A 533-B and A 508 by western manufacturers. (author)

  4. Pressure vessel steels: influence of chemical composition on irradiation sensitivity

    International Nuclear Information System (INIS)

    Ghoniem, M.M.; Hammad, F.H.

    1998-01-01

    Neutron irradiation of the steels used in the construction of the nuclear reactor pressure vessels can lead to the embrittlement of these materials, increasing the ductile-to-brittle transition temperature and decreasing the fracture energy, which can limit the plant life. The knowledge of irradiation embrittlement and the means for minimizing such degradation is therefore important in the field of assuring the safety of the nuclear power plants. Irradiation embrittlement is quite a complex process. It involves many variables. The most important of these are irradiation temperature, neutron fluence (neutron dose), neutron flux (neutron dose rate), and chemical composition of the irradiated material. This paper is concerned with the effect of chemical composition, the role of residual and alloying elements in the irradiation embrittlement of nuclear reactor pressure vessel steels in light water reactors. It presents a critical review for the published work in this field through the last 25 years

  5. Studies of fragileness in steels of vessels of BWR reactors

    International Nuclear Information System (INIS)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2003-01-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA MARK lll reactor and separately with Ni +3 ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A 2 . (Author)

  6. 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

  7. Development of the plant life management technology for RPV steels [ - Current status of surveillance test specimen reconstitution program -

    International Nuclear Information System (INIS)

    Kazunobu, Sakamoto; Eliichiro, Otsuka; Yoshiaki, Oka; Kosei, Taguchi; Michiyoshi, Yamamoto

    2001-01-01

    In order to develop the reconstitution technology to standardize surveillance test specimen reconstitution practices to deal with the extended service life of reactor pressure vessels, the Japan Power Engineering and Inspection Corporation (JAPEIC) has been carried out the project entrusted by the Ministry of International Trade and Industry (MITI). We focus on a correlation between the reduction of absorbed energy and the interaction of the heat affected zone (HAZ) and the plastic zone, to establish applicable reconstitution conditions for Charpy specimens. The relationship between the plastic zone width and the absorbed energy has been obtained by estimating the plastic zone width from the hardness distribution of the Charpy specimens. Impact tests of reconstituted specimens with 10 mm-length insert using the surface activated joining method were performed and the test results were compared to those obtained by the standard specimens. By comparing the length of insert material to the sum of HAZ width and plastic zone width, it is clear that the interaction causes the reduction of the absorbed energy. Hence, the applicable conditions of reconstituted Charpy specimens could be assessed by comparing the insert length to the sum of HAZ width and plastic zone width. Moreover the effects of the possible deviations from the standard shape and size specimens for the reconstituted specimens were studied. (authors)

  8. 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.

  9. Corrosion fatigue of pressure vessel steels in PWR environments--influence of steel sulfur content

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.M.; Druce, S.G.; Truswell, A.E.

    1984-07-01

    Large effects of simulated light water reactor environments at 288 C on fatigue crack growth in low alloy pressure vessel steels are observed only when specific mechanical, metallurgical, and electrochemical conditions are satisfied simultaneously. In this paper, the relative importance of three key variables--steel impurity content, water chemistry, and flow rate--and their interaction with loading rate or strain rate are examined. In particular, the results of a systematic examination of the influence of a steel's sulfur content are described.

  10. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Roychowdhury, S.; Seifert, H.-P.; Spätig, P.; Que, Z.

    2016-01-01

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  11. Effect of high-temperature water and hydrogen on the fracture behavior of a low-alloy reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Roychowdhury, S., E-mail: sroy27@gmail.com [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland); Materials Processing & Corrosion Engineering Division, Mod-Lab, D-Block, Bhabha Atomic Research Centre, Mumbai 400085 (India); Seifert, H.-P.; Spätig, P.; Que, Z. [Paul Scherrer Institut, Nuclear Energy and Safety Research Department, Laboratory for Nuclear Materials, 5232 Villigen, PSI (Switzerland)

    2016-09-15

    Structural integrity of reactor pressure vessels (RPV) is critical for safety and lifetime. Possible degradation of fracture resistance of RPV steel due to exposure to coolant and hydrogen is a concern. In this study tensile and elastic-plastic fracture mechanics (EPFM) tests in air (hydrogen pre-charged) and EFPM tests in hydrogenated/oxygenated high-temperature water (HTW) was done, using a low-alloy RPV steel. 2–5 wppm hydrogen caused embrittlement in air tensile tests at room temperature (25 °C) and at 288 °C, effects being more significant at 25 °C and in simulated weld coarse grain heat affected zone material. Embrittlement at 288 °C is strain rate dependent and is due to localized plastic deformation. Hydrogen pre-charging/HTW exposure did not deteriorate the fracture resistance at 288 °C in base metal, for investigated loading rate range. Clear change in fracture morphology and deformation structures was observed, similar to that after air tests with hydrogen. - Highlights: • Hydrogen content, microstructure of LAS, and strain rate affects tensile properties at 288 °C. • Strength affects hydrogen embrittlement susceptibility to a greater extent than grain size. • Hydrogen in LAS leads to strain localization and restricts cross-slip at 288 °C. • Possible hydrogen pickup due to exposure to 288 °C water alters fracture surface appearance without affecting fracture toughness in bainitic base material. • Simulated weld heat affected zone microstructure shows unstable crack propagation in 288 °C water.

  12. A study of the mechanisms for the irradiation embrittlement of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Solt, G.; Zimmermann, U.; Waeber, W.B.; Mercier, O.; Frisius, F.; Ghazi-Wakili, K.

    1987-03-01

    Irradiation damage particles were detected by small angle neutron scattering and positron annihilation techniques in two RPV steels. The particle radii were 8A and 14A prior to heat treatments for the plate and weldment, respectively; annealing leads to coarsening in the weldment, the volume fraction remains essentially constant at about 0.14%. The model of copper-rich precipitates 'diluted' by Mn atoms or, alternatively, by vacancy agglomerates is consistent with the neutron scattering data, the presence of simple voids in the weldment would contradict the positron results. Preliminary results on these steels and also on related alloys by methods 'new' in this field are reported. (author)

  13. 46 CFR 42.09-30 - Additional survey requirements for steel-hull vessels.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Additional survey requirements for steel-hull vessels...-30 Additional survey requirements for steel-hull vessels. (a) In addition to the requirements in § 42...) When the vessel is in drydock, the hull plating, etc., shall be examined. (c) The holds, 'tween decks...

  14. Effect of tempering temperature on the microstructure and mechanical properties of a reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Li, C.W.; Han, L.Z.; Luo, X.M.; Liu, Q.D.; Gu, J.F., E-mail: gujf@sjtu.edu.cn

    2016-08-15

    The microstructure and mechanical properties of reactor pressure vessel (RPV) steel were investigated after tempering at different temperatures ranging from 580 to 700 °C for 5 h. With increasing tempering temperature, the impact toughness, which is qualified by Charpy V-notch total absorbed energy, initially increases from 142 to 252 J, and then decreases to 47 J, with a maximum value at 650 °C, while the ultimate tensile strength varies in exactly the opposite direction. Comparing the microstructure and fracture surfaces of different specimens, the variations in toughness and strength with the tempering temperature were generally attributed to the softening of the bainitic ferrite, the agminated Fe{sub 3}C carbides that resulted from decomposition of martensite/austenite (M/A) constituents, the precipitation of Mo{sub 2}C carbides, and the newly formed M/A constituents at the grain boundaries. Finally, the correlation between the impact toughness and the volume fraction of the M/A constituents was established, and the fracture mechanisms for the different tempering conditions are explained. - Highlights: • The dependence of the deterioration of impact toughness on tempering temperature has been analysed. • The instrumented Charpy V-notch impact test has been employed to study the fracture mechanism. • The influence of M/A constituents on different fracture mechanisms based on the hinge model has been demonstrated. • A correlation between the mechanical properties and the amount of M/A constituents has been established.

  15. Study of the flux effect nature for VVER-1000 RPV welds with high nickel content

    Energy Technology Data Exchange (ETDEWEB)

    Kuleshova, E.A., E-mail: evgenia-orm@yandex.ru [National Research Center “Kurchatov Institute”, Kurchatov Sq.1, 123182, Moscow (Russian Federation); National Research Nuclear University, “MEPhI” (Moscow Engineering Physics Institute), Kashirskoe Highway 31, 115409, Moscow (Russian Federation); Gurovich, B.A.; Lavrukhina, Z.V.; Maltsev, D.A.; Fedotova, S.V.; Frolov, A.S.; Zhuchkov, G.M. [National Research Center “Kurchatov Institute”, Kurchatov Sq.1, 123182, Moscow (Russian Federation)

    2017-01-15

    This work extends the research of the basic regularities of segregation processes in the grain boundaries (GB) of VVER-1000 reactor pressure vessel (RPV) steels. The paper considers the influence of irradiation with different fast neutron fluxes on the structure, yield strength and ductile-to-brittle transition temperature (T{sub K}) changes as well as on changes of the share of brittle intergranular fracture and development of segregation processes in the VVER-1000 RPV weld metal (WM). The obtained experimental results allow to separate the contribution of the hardening and non-hardening mechanisms to mechanical properties degradation of material irradiated at the operating temperature. It is shown that the difference in T{sub K} shift in WM irradiated to the same fluence with different fast neutron fluxes is mainly due to the difference in the GB accumulation kinetics of impurities and only to a small extent due to the material hardening. Phosphorus bulk diffusion coefficients were evaluated for the temperature exposure, accelerated irradiation and irradiation within surveillance specimens (SS) using a kinetic model of phosphorus GB accumulation in low-alloyed low-carbon steels under the influence of operational factors. The correlation between the GB segregation level of phosphorus and nickel, and the T{sub K} shift - in WM SS was obtained experimentally and indicates the non-hardening mechanism contribution to the total radiation embrittlement of VVER-1000 RPV steels throughout its extended lifetime. - Highlights: • Structural elements in high Ni welds are studied at different irradiation fluxes. • AES study demonstrated different P GB kinetics at different irradiation fluxes. • Hardening and non-hardening mechanism contributions to the flux effect are assessed. • Correlation between ΔT{sub K} and P and Ni GB content is shown for VVER-1000 RPV welds.

  16. Evaluation of ductile-brittle transition behavior with neutron irradiation in nuclear reactor pressure vessel steels using small punch test

    International Nuclear Information System (INIS)

    Kim, M. C.; Lee, B. S.; Oh, Y. J.

    2003-01-01

    A Small Punch (SP) test was performed to evaluate the ductile-brittle transition temperature before and after neutron irradiation in Reactor Pressure Vessel (RPV) steels produced by different manufacturing (refining) processes. The results were compared to the standard transition temperature shifts from the Charpy test and Master Curve fracture toughness test in accordance with the ASTM standard E1921. The samples were taken from 1/4t location of the vessel thickness and machined into a 10x10x0.5mm dimension. Irradiation of the samples was carried out in the research reactor at KAERI (HANARO) at about 290 .deg. C of the different fluence levels respectively. SP tests were performed in the temperature range of RT to -196 .deg. C using a 2.4mm diameter ball. For the materials before and after irradiation, SP transition temperatures (T sp ), which are determined at the middle of the upper and lower SP energies, showed a linear correlation with the Charpy index temperature, T 41J . T sp from the irradiated samples was increased as the fluence level increased and was well within the deviation range of the unirradiated data. The TSP had a correlation with the reference temperature (T 0 ) from the master curve method using a pre-cracked Charpy V-notched (PCVN) specimen

  17. Nuclear reactors sited deep underground in steel containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bourque, Robert [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2006-07-01

    Although nuclear power plants are certainly very safe, they are not perceived as safe by the general populace. Also, there are concerns about overland transport of spent fuel rods and other irradiated components. It is hereby proposed that the nuclear components of nuclear power plants be placed in deep underground steel vessels with secondary coolant fed from them to turbines at or near the surface. All irradiated components, including spent fuel, would remain in the chamber indefinitely. This general concept was suggested by the late Edward Teller, generated some activity 20-25 years ago and appears to be recently reviving in interest. Previous work dealt with issues of geologic stability of underground, possibly reinforced, caverns. This paper presents another approach that makes siting independent of geology by placing the reactor components in a robust steel vessel capable of resisting full overburden pressure as well as pressures resulting from accident scenarios. Structural analysis of the two vessel concepts and approximate estimated costs are presented. This work clears the way for the extensive discussions required to evaluate the advantages of this concept. (author)

  18. Analysis of the procedure proposed by AREVA to prove adequate toughness of the domes of the Flamanville 3 EPR reactor pressure vessel (RPV) lower head and closure head. Session of 30 September 2015. Public version

    International Nuclear Information System (INIS)

    Catteau, R.; Cadet-Mercier, S.

    2015-01-01

    AREVA has asked ASN to evaluate the conformity of the reactor pressure vessel (RPV) for the Flamanville 3 EPR in application of the order reference [6]. The domes of the Flamanville 3 RPV closure head and lower head were manufactured in 2006 and 2007. AREVA identified that these components displayed a risk of heterogeneity of their characteristics and therefore carried out a technical qualification. At the end of 2014, AREVA informed ASN of lower-than-expected results of impact tests conducted as part of this technical qualification on test specimens taken from a dome representative of those intended for Flamanville 3. The values measured on two series of three test specimens give a mean value of 52 joules which does not attain the quality standard expected by AREVA. This mean value is also lower than the bending rupture energy value of 60 joules mentioned in point 4 of appendix 1 of the order reference [6], with which compliance would have been sufficient to prove the toughness of the material. AREVA carried out investigations to determine the origin of these noncompliant values. The carbon concentration measurements taken at the surface of the representative dome by portable spectrometry revealed the presence of a zone of major positive segregation (high concentration of carbon) over a diameter of about one meter. Furthermore, the examinations show that the segregation extends to a depth exceeding a quarter of the thickness of the dome. AREVA explains the non-compliance with the bending rupture energy criterion by the presence of this major positive segregation which came from the ingot used for the forging and was not completely eliminated by the cropping operations. To deal with this deviation, AREVA plans proving that the material is sufficiently tough by conducting new tests on a material that is representative of the lower and upper domes of the Flamanville EPR reactor. The body of the Flamanville 3 RPV, of which the lower dome is a part, has already

  19. Preliminary results of steel containment vessel model test

    International Nuclear Information System (INIS)

    Matsumoto, T.; Komine, K.; Arai, S.

    1997-01-01

    A high pressure test of a mixed-scaled model (1:10 in geometry and 1:4 in shell thickness) of a steel containment vessel (SCV), representing an improved boiling water reactor (BWR) Mark II containment, was conducted on December 11-12, 1996 at Sandia National Laboratories. This paper describes the preliminary results of the high pressure test. In addition, the preliminary post-test measurement data and the preliminary comparison of test data with pretest analysis predictions are also presented

  20. 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

  1. 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)

  2. Dynamic fracture characterization of a pressure vessel steel

    International Nuclear Information System (INIS)

    Schmitt, W.; Boehme, W.; Klemm, W.; Memhard, D.; Winkler, S.

    1991-01-01

    Dynamic events are characterized by time and space-dependent stress and strain fields caused by wave or inertia effect. The dynamic effect at cracks may be originated from the rapid loading rate or impact loading of a structure containing a stationary crack or the time-dependent stress and strain fields of a propagating or arresting crack itself. Dynamic effects complicate the analysis of crack tip stress and strain fields, and usually considerable experimental effort and numerical technique are required. High loading rate influences the deformation and yield behavior and also the fracture toughness of materials. In order to know the propagation and arrest behavior of cracks, a heat of a German reactor pressure vessel steel was investigated, and the dynamic J-resistance curves were evaluated with large three-point bending specimens by impact loading, moreover, the crack propagation energy at large crack extension was determined with wide tension plates. The material tested was a ferritic pressure vessel steel, ASTM A 508 Cl 2. The dynamic J-resistance curves and numerical simulation and fractographic examination, and crack propagation energy are reported. (K.I.)

  3. Analytical modeling of the effect of crack depth, specimen size, and biaxial stress on the fracture toughness of reactor vessel steels

    International Nuclear Information System (INIS)

    Chao, Yuh-Jin

    1995-01-01

    Fracture, toughness values for A533-B reactor pressure vessel (RPV) steel obtained from test programs at Oak Ridge National Laboratory (ORNL) and University of Kansas (KU) are interpreted using the J-A 2 analytical model. The analytical model is based on the critical stress concept and takes into consideration the constraint effect using the second parameter A 2 in addition to the generally accepted first parameter J which represents the loading level. It is demonstrated that with the constraint level included in the model effects of crack depth (shallow vs deep), specimen size (small vs. large), and loading type (uniaxial vs biaxial) on the fracture toughness from the test programs can be interpreted and predicted

  4. The near-threshold high R-ratio fatigue crack growth characteristics of SA508 cl III reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Achilles, R.D.; Bulloch, J.H.

    1989-01-01

    This paper describes the effect of frequency and environment on the near-threshold fatigue crack growth behaviour of SA508 cl III reactor pressure vessel (RPV) steel. The study has shown that in the near-threshold regime microstructure and environment markedly affect fatigue crack growth behaviour. In an aqueous environment, fatigue crack growth behaviour became even more sensitive to microstructure, and the fatigue crack growth rate increased by a factor of four in the case of the 3Hz test, while that for the 0.3Hz test was increased by a factor of approximately sixteen. This environmental enhancement manifested itself in the form of intergranular failure. For the 0.3Hz test the percentage intergranular failure decreased from 18% to <1% with an increase in ΔK level. The transition from microstructure-sensitive to microstructure-insensitive occurs when the cyclic plastic zone size is of the order of the prior austenite grain size. (author)

  5. Residual stresses in weld-clad reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Bertram, W.

    1975-01-01

    Cladding of low alloy nuclear reactor pressure vessel steel with austenitic stainless steel introduces in heavy section components high residual stresses which may cause microcrack formation in stress relief heat treatment. In this investigation an attempt is made to contribute to the solution of the stress relief cracking problem by determining quantitatively the magnitude and distribution of the residual stresses after cladding and after subsequent stress relief heat treatment. The distribution of residual stresses was determined on the basis of a combined experimental-mathematical procedure. Heavy section plate specimens of low alloy steel as base material were given an austenitic monolayer-cladding using the techniques of strip electrode and plasma hot wire cladding, respectively. A number of plates was stress relief heat treated. Starting from the cladded surface the thickness of the plates was reduced by subsequent removal of layers of material. The elastic strain reaction to the removal of each layer was measured by strain gauges. From the data obtained the biaxial residual stress distribution was computed as a function of thickness using relations which are derived for this particular case. In summary, lower residual stresses are caused by reduced thickness of the components. As the heat input, is decreased at identical base material thickness, the residual stresses are lowered also. The height of the tensile residual stress peak, however, remains approximataly constant. In stress relief annealed condition the residual stresses in the cladding are in tension; in the base material the residual stresses are negligibly small

  6. 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

  7. Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

    International Nuclear Information System (INIS)

    Loktionov, V.D.; Mukhtarov, E.S.; Yaroshenko, N.I.; Orlov, V.E.

    1999-01-01

    Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of 'overpressure-molten pool structure' when the vessel failure started at the 'hot' layers of the vessel. (orig.)

  8. Irradiation and annealing behavior of 15Kh2MFA reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Popp, K.; Bergmann, U.; Bergner, F.; Hampe, E.; Leonhardt, W.D.; Schuetzler, H.P.; Viehrig, H.W.

    1992-01-01

    This work deals with the mechanical properties of RPV steels used WWER-440. The materials under investigation were a forging (base metal 15Kh2MFA) and the corresponding weld. Charpy V-notch specimens and tensile test specimens were irradiated in the WWER-2 Rheinsberg at about 270 C up to the two neutron fluence levels of 4 x 10 18 and 5 x 10 19 n/cm 2 (E>1MeV). Post-irradiation annealing heat treatments were performed, among others a 475 C/152 h treatment of technical interest. (orig.)

  9. Effect of Loading Transients on the EAC Crack Growth Behaviour of Low-Alloy RPV Steels under Simulated BWR Operating Conditions (CASTOC WP 3, PSI Tests 3 and 4)

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.-P

    2003-04-01

    Within the CASTOC-project (5{sup t}h EU FW programme), the environmentally-assisted crack (EAC) growth behaviour of low-alloy reactor pressure vessel (RPV) steels is experimentally investigated under simulated transient and steady-state boiling water reactor (BWR) power operation conditions by six European laboratories. The present report is a summary of the third and fourth test of working package (WP) 3 with loading transients, performed at Paul Scherrer Institut (PSI). Two different low-alloy steels (20 MnMoNi 5 5, 0.015 wt.% S and 22 NiMoCr 3 7, 0.007 wt. %S) were investigated in oxygenated high-temperature, high-purity water (T = 240 {sup o}C, DO = 400 ppb) in a daisy chain at two different load ratios (R = 0.8 and 0.2). In the first part of the experiments, asymmetrical saw tooth loading with different rise times {delta}t{sub R} of the load and different loading frequencies were applied. Then the loading conditions were changed to an asymmetrical trapezoid waveform loading (periodical partial unloading, PPU) and the hold time {delta}t{sub H} at maximum load was varied. In the final phase of WP 3 PSI tests 3 and 4 the SCC behaviour was investigated under constant load. With decreasing loading frequency the corrosion fatigue (CF) crack advance per cycle {delta}a/{delta}N{sub EAC} of material A increased. Sustained EAC crack growth could be maintained down to low frequencies of 10{sup -5} Hz. The time-based crack growth rate (CGR) da/dt{sub EAC} decreased with decreasing frequency. In material B no effect of the loading frequency could be resolved. Up to a hold time of 1 h at maximum constant load the CGR da/dt{sub EAC} seemed to be independent of the hold time. Above hold times of 1 h the CGR decreased and dropped down to CGR values in the range or below the BWR VIP 60 SCC disposition lines. This behaviour was observed in both investigated materials. The cycle-based CGR {delta}a/{delta}N{sub EAC} remained approximately constant with increasing hold time. The

  10. Effect of aging on properties of pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Druce, S.G.; Gage, G.; Jordan, G.

    1986-04-01

    Manganese-molybdenum-nickel steels are used in nuclear pressure vessels operating at temperatures up to 350/sup 0/C. The effects of thermal ageing in the temperature range 300-550/sup 0/C for durations up to 2 x 10/sup 4/ h have been studied in conventionally quenched and tempered and simulated heat-affected-zone (HAZ) microstructural conditions. Quantitative fractography and Auger spectroscopy have been used to relate changes in mechanical properties with changes in fracture mode and grain boundary chemistry. Aging increases the ductile-brittle transition temperature by an amount dependent on material, prior heat treatment, aging temperature and time. Embrittlement is associated with segregation of phosphorus to grain boundaries and is modelled using McLean's approach to equilibrium segregation.

  11. 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

  12. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  13. Research to sustain cases for Magnox-reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Graham, W.J.

    1997-01-01

    Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

  14. Normalizing treatment influence on the forged steel SAE 8620 fracture properties

    Directory of Open Access Journals (Sweden)

    Paulo de Tarso Vida Gomes

    2005-03-01

    Full Text Available In a PWR nuclear power plant, the reactor pressure vessel (RPV contains the fuel assemblies and reactor vessels internals and keeps the coolant at high temperature and high pressure during normal operation. The RPV integrity must be assured all along its useful life to protect the general public against a significant radiation liberation damage. One of the critical issues relative to the VPR structural integrity refers to the pressurized thermal shock (PTS accident evaluation. To better understand the effects of this kind of event, a PTS experiment has been planned using an RPV prototype. The RPV material fracture behavior characterization in the ductile-brittle transition region represents one of the most important aspects of the structural assessment process of RPV's under PTS. This work presents the results of fracture toughness tests carried out to characterize the RPV prototype material behavior. The test data includes Charpy energy curves, T0 reference temperatures for definition of master curves, and fracture surfaces observed in electronic microscope. The results are given for the vessel steel in the "as received" and normalized conditions. This way, the influence of the normalizing treatment on the fracture properties of the steel could be evaluated.

  15. 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.

  16. Microscopic examination of crack growth in a pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Isacsson, M.; Narstroem, T. [Royal Inst. of Tech., Stockholm (Sweden)

    1997-01-01

    A fairly systematic microscopic study concerning ductile and ductile-brittle crack growth in the A508B pressure vessel steel has been performed. The main method of investigation was to subject fracture mechanics specimens (sub-sized three point bend specimens) to predetermined load levels corresponding to different amounts of ductile crack extension. The specimens were then cut perpendicularly to the plane of the crack and the area in front of the crack was examined in a SEM. The object of these examinations was to determine if newly encountered computational results could be correlated to crack extension characteristics and to study whether the mechanism of ductile growth was of the void growth type or of the fast shear mechanism. This is important for further numerical modelling of the process. Both the original material and a specially heat treated piece were investigated. The heat treatment was performed in order to raise the transition temperature to about 60 deg C with the object to provide a more convenient testing situation. Charpy V tests were performed for the specially heat treated material to obtain the temperature dependence of the toughness. This was also studied by performing fracture toughness determination on the same type of specimens as were used for the microscopic study. The heat treatment used fulfilled the above purpose and the microscopic studies provide a good understanding of the micro mechanisms operating in the ductile fracture process for this material. 19 refs, 8 figs, 3 tabs.

  17. Microscopic examination of crack growth in a pressure vessel steel

    International Nuclear Information System (INIS)

    Isacsson, M.; Narstroem, T.

    1997-01-01

    A fairly systematic microscopic study concerning ductile and ductile-brittle crack growth in the A508B pressure vessel steel has been performed. The main method of investigation was to subject fracture mechanics specimens (sub-sized three point bend specimens) to predetermined load levels corresponding to different amounts of ductile crack extension. The specimens were then cut perpendicularly to the plane of the crack and the area in front of the crack was examined in a SEM. The object of these examinations was to determine if newly encountered computational results could be correlated to crack extension characteristics and to study whether the mechanism of ductile growth was of the void growth type or of the fast shear mechanism. This is important for further numerical modelling of the process. Both the original material and a specially heat treated piece were investigated. The heat treatment was performed in order to raise the transition temperature to about 60 deg C with the object to provide a more convenient testing situation. Charpy V tests were performed for the specially heat treated material to obtain the temperature dependence of the toughness. This was also studied by performing fracture toughness determination on the same type of specimens as were used for the microscopic study. The heat treatment used fulfilled the above purpose and the microscopic studies provide a good understanding of the micro mechanisms operating in the ductile fracture process for this material

  18. Kinetics of annealing of irradiated surveillance pressure vessel steel

    International Nuclear Information System (INIS)

    Harvey, D.J.; Wechsler, M.S.

    1982-01-01

    Indentation hardness measurements as a function of annealing were made on broken halves of Charpy impact surveillance samples. The samples had been irradiated in commercial power reactors to a neutron fluence of approximately 1 x 10 18 neutrons per cm 2 , E > 1 MeV, at a temperature of about 300 0 C (570 0 F). Results are reported for the weld metal, which showed greater radiation hardening than the base plate or heat-affected zone material. Isochronal and isothermal anneals were conducted on the irradiated surveillance samples and on unirradiated control samples. No hardness changes upon annealing occurred for the control samples. The recovery in hardness for the irradiated samples took place mostly between 400 and 500 0 C. Based on the Meechan-Brinkman method of analysis, the activation energy for annealing was found to be 0.60 +- 0.06 eV. According to computer simulation calculations of Beeler, the activation energy for migration of vacancies in alpha iron is about 0.67 eV. Therefore, the results of this preliminary study appear to be consistent with a mechanism of annealing of radiation damage in pressure vessel steels based on the migration of radiation-produced lattice vacancies

  19. Hydraulics in the RPV lower-plenum of EPR

    International Nuclear Information System (INIS)

    Barois, G.; Goreaud, N.; Nicaise, N.

    2001-01-01

    The in-core instrumentation penetrations of the European Pressurised water Reactor (EPR) have been removed from RPV-bottom to RPV-head, leaving empty the lower plenum of the RPV (Reactor Pressure Vessel). In a lower plenum with no internal structure, huge vortices may appear, with negative consequences, such as high disturbance of the core inlet flow distribution, and high increase of the RPV pressure loss. FRAMATOME ANP developed a specific Flow Distribution Device (FDD), annular shaped, located in the RPV lower plenum below the core support plate, which prevents huge vortices from appearing and guarantees a satisfying flow distribution at core inlet in normal operating conditions. The design of the FDD has been optimised with a numerical approach, using the 3-D CFD-code STAR-CD, previously qualified on scale mockup tests. The model developed represents the EPR RPV from the cold leg to core inlet. Thus, the flow distribution at core inlet, the mixing between loop-flows upstream core inlet and the pressure loss in the lower plenum can be evaluated. The optimised FDD provides satisfying performances for all these relevant functional items. (author)

  20. 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

  1. 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)

  2. Estimation of residual stresses in reactor pressure vessel steel specimens clad by stainless steel strip electrodes

    International Nuclear Information System (INIS)

    Schimmoeller, H.A.; Ruge, J.L.

    1978-01-01

    The equations to determine a two-dimensional state of residual stress in flat laminated plates are well known from an earlier work by one of the authors. The derivation of these equations leads to a linear, inhomogeneous system of Volterra's integral equations of the second kind. To ascertain the unknown residual stresses from these equations it is necessary to cut down the thickness of the test plate layer by layer. This results in two-dimensional deformation reactions in the rest of the test plate, which can be measured, e.g. by a strain gauge rosette applied to the opposite side of the plate. The above-mentioned stress analysis has been transferred to 86mm thick reactor pressure vessel steel specimens (Type 22NiMoCr 37, DIN-No. 1.6751, similar to ASTM A508, Class 2) double-run clad by austenitic stainless steel strip electrodes (first layer 24/13 Cr-Ni steel, second layer 21/10 Cr-Ni steel). The overall dimensions of the clad specimens investigated amounted to 200 x 200 x (86+4.5+4.5)mm. At the surface of the austenitic cladding there is a two-dimensional tensile normal stress state of about 200N/mm 2 parallel, and about 300N/mm 2 transverse, to the welding direction. The maximum tensile stress was 8mm below the interface (fusion line, material transition) in the parent material. The stress distributions of the specimens investigated, determined on the basis of the above-mentioned combined experimental mathematical procedure, are presented graphically for the as-welded (as-delivered) and annealed (600 0 C/12hr) conditions. (author)

  3. Performance of core modifications to reduce the reactor pressure vessel fluence

    International Nuclear Information System (INIS)

    Kiehlmann, H.D.; Lisdat, R.; Sommer, D.

    1997-01-01

    It's often discussed that nuclear power plants (NPP) are designed for an operation of 40 years equivalent to 32 full power years (FPY) assuming a load factor of 0.8. Such fixed plant life times are subjects of US operating licenses but not, as in most other countries, in the Federal Republic of Germany. Here the operating licenses are issued for an indefinite period. However, the German utilities are continuously upgrading their plants to attain a safety level that meets all current requirements. These upgrading measures also include the replacement of bigger components like e.g. the steam generator. The reactor pressure vessel (RPV), however, has a special status. Unlike most other components of a NPP which most likely will be exchanged during its service life a replacement or annealing treatment of the RPV certainly require more efforts to be economically justified. Thus the embrittlement of the RPV has an essential impact on the life time of a NPP. The end-of-life (EOL) RPV material toughness in essential depends on the steel quality and the accumulated neutron fluence. For a given NPP the reduction of the neutron flux at the inner surface of the RPV is the only way to limit its embrittlement. The resulting modifications for the core loadings in combination with the insertion of additional core components like steel elements are described and the impact on core performance and RPV fluence considered. (UK)

  4. Study on ex-vessel cooling of RPV (behavior of coalesced bubbles and trigger condition of critical heat flux on inclined plate)

    International Nuclear Information System (INIS)

    Ohtake, H.; Koizumi, Y.; Takano, K.I.

    2001-01-01

    The Ex-vessel cooling of Reactor-Pressure-Vessel in Light-Water-Reactor at the severe accident have been proposed for future nuclear reactors. The estimation of Critical-Heat-Flux on a downward-facing curvilinear surface, like a hemisphere, is important to the assessment of the cooling. In this study, the CHFs on inclined surfaces were examined experimentally focusing on orientation of the heating surface. In order to discuss detailed mechanism of the CHF, the behaviors of coalesced bubbles near the heating surface were investigated through visual observations. The critical heat flux obtained in the present experiments increased with the inclined angle over the present experimental range. The dependence of the inclined angle on the critical heat flux was q CHF,R-113 [q] = f (q 0.33 ) for the present experimental results. The effect of the surface orientation on the critical heat flux was roughly explained by using the simple analytical model based on the macro-layer model and Kelvin-Helmholtz instability. From visual observations for behavior of bubbles near the heating surface, whereas the coalesced bubble covered over the heating surface for the inclined angle of 0 degree, the coalesced bubble moved upward to avoid packing the bubble on the surface above 5 degree. As the inclined angle increased, the velocity of the coalesced bubble was high, the period covered the heater and the bubble length were small. The results suggested that the CHF was closely related to forming the coalesced bubble and the behavior of the bubble. (author)

  5. Experimental data base for assessment of irradiation induced ageing effects in pre-irradiated RPV materials of German PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hein, H.; Gundermann, A.; Keim, E.; Schnabel, H. [AREVA NP GmbH (Germany); Ganswind, J. [VGB PowerTech e.V (Germany)

    2011-07-01

    The 5 year research program CARISMA which ended in 2008 has produced a data base to characterize the fracture toughness of pre-irradiated original RPV (Reactor Pressure Vessel) materials being representative for all four German PWR construction lines of former Siemens/KWU company. For this purpose tensile, Charpy-V impact, crack initiation and crack arrest tests have been performed for three base materials and four weld metals irradiated to neutron fluences beyond the designed EoL range. RPV steels with optimized chemical composition and with high copper as well as high nickel content were examined in this study. The RTNDT concept and the Master Curve approach were applied for the assessment of the generated data in order to compare both approaches. A further objective was to clarify in which extent crack arrest curves can be generated for irradiated materials and how crack arrest can be integrated into the Master Curve approach. By the ongoing follow-up project CARINA the experimental data base will be extended by additional representative materials irradiated under different conditions and with respect to the accumulated neutron fluences and specific impact parameters such as neutron flux and manufacturing effects. The irradiation data cover also the long term irradiation behavior of the RPV steels concerned. Moreover, most of the irradiated materials were and will be used for microstructural examinations to get a deeper insight in the irradiation embrittlement mechanisms and their causal relationship to the material property changes. By evaluation of the data base the applicability of the Master Curve approach for both crack initiation and arrest was confirmed to a large extent. Moreover, within both research programs progress was made in the development of crack arrest test techniques and in specific issues of RPV integrity assessment. (authors)

  6. The electrogas and electroslag multipass high speed welding of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Eichhorn, F.; Hirsch, P.; Langenbahn, H.W.; Wubbels, B.

    1978-01-01

    High-speed electroslag and electrogas welding of 15 Mn Ni63 steel plates to achieve high strength and toughness joints for reactor pressure vessels are described. Mechanical testing of overheating-resistant, brittle fracture resistant low alloy steels is discussed. (UK)

  7. 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)

  8. Nonlinear Ultrasonic Techniques to Monitor Radiation Damage in RPV and Internal Components

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, Laurence [Georgia Inst. of Technology, Atlanta, GA (United States); Kim, Jin-Yeon [Georgia Inst. of Technology, Atlanta, GA (United States); Qu, Jisnmin [Northwestern Univ., Evanston, IL (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wall, Joe [Electric Power Research Inst. (EPRI), Knoxville, TN (United States)

    2015-11-02

    The objective of this research is to demonstrate that nonlinear ultrasonics (NLU) can be used to directly and quantitatively measure the remaining life in radiation damaged reactor pressure vessel (RPV) and internal components. Specific damage types to be monitored are irradiation embrittlement and irradiation assisted stress corrosion cracking (IASCC). Our vision is to develop a technique that allows operators to assess damage by making a limited number of NLU measurements in strategically selected critical reactor components during regularly scheduled outages. This measured data can then be used to determine the current condition of these key components, from which remaining useful life can be predicted. Methods to unambiguously characterize radiation related damage in reactor internals and RPVs remain elusive. NLU technology has demonstrated great potential to be used as a material sensor – a sensor that can continuously monitor a material’s damage state. The physical effect being monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave. The degree of nonlinearity is quantified with the acoustic nonlinearity parameter, β, which is an absolute, measurable material constant. Recent research has demonstrated that nonlinear ultrasound can be used to characterize material state and changes in microscale characteristics such as internal stress states, precipitate formation and dislocation densities. Radiation damage reduces the fracture toughness of RPV steels and internals, and can leave them susceptible to IASCC, which may in turn limit the lifetimes of some operating reactors. The ability to characterize radiation damage in the RPV and internals will enable nuclear operators to set operation time thresholds for vessels and prescribe and schedule replacement activities for core internals. Such a capability will allow a more clear definition of reactor safety margins. The research consists of three tasks: (1

  9. Nonlinear Ultrasonic Techniques to Monitor Radiation Damage in RPV and Internal Components

    International Nuclear Information System (INIS)

    Jacobs, Laurence; Kim, Jin-Yeon; Qu, Jisnmin; Ramuhalli, Pradeep; Wall, Joe

    2015-01-01

    The objective of this research is to demonstrate that nonlinear ultrasonics (NLU) can be used to directly and quantitatively measure the remaining life in radiation damaged reactor pressure vessel (RPV) and internal components. Specific damage types to be monitored are irradiation embrittlement and irradiation assisted stress corrosion cracking (IASCC). Our vision is to develop a technique that allows operators to assess damage by making a limited number of NLU measurements in strategically selected critical reactor components during regularly scheduled outages. This measured data can then be used to determine the current condition of these key components, from which remaining useful life can be predicted. Methods to unambiguously characterize radiation related damage in reactor internals and RPVs remain elusive. NLU technology has demonstrated great potential to be used as a material sensor - a sensor that can continuously monitor a material's damage state. The physical effect being monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave. The degree of nonlinearity is quantified with the acoustic nonlinearity parameter, β, which is an absolute, measurable material constant. Recent research has demonstrated that nonlinear ultrasound can be used to characterize material state and changes in microscale characteristics such as internal stress states, precipitate formation and dislocation densities. Radiation damage reduces the fracture toughness of RPV steels and internals, and can leave them susceptible to IASCC, which may in turn limit the lifetimes of some operating reactors. The ability to characterize radiation damage in the RPV and internals will enable nuclear operators to set operation time thresholds for vessels and prescribe and schedule replacement activities for core internals. Such a capability will allow a more clear definition of reactor safety margins. The research consists of three tasks

  10. Irradiation effects in low-alloy reactor pressure vessel steels (Heavy-Section Steel Technology program series 4 and 5)

    International Nuclear Information System (INIS)

    McGowan, J.J.; Nanstad, R.K.; Thoms, K.R.; Menke, B.H.

    1985-01-01

    This report presents studies on the irradiation effects in low-alloy reactor pressure vessel steels. The Fourth Heavy-Section Steel Technology (HSST) Irradiation Series, almost completed, was aimed at elastic-plastic and fully plastic fracture toughness of low-copper weldments (''current practice welds''). A typical nuclear pressure vessel plate steel was included for statistical purposes. The Fifth HSST Irradiation Series, now in progress, is aimed at determining the shape of the K/sub IR/ curve after significant radiation-induced shift of the transition temperatures. This series includes irradiated test specimens of thicknesses up to 100 mm and weldment compositions typical of early nuclear power reactor pressure vessel welds. 27 refs., 22 figs

  11. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1994-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (74-90 mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments. Under severe accident loading conditions, the steel containment vessel in a typical Mark-I or Mark-II plant may deform under internal pressurization such that it contacts the inner surface of a shield building wall. (Thermal expansion from increasing accident temperatures would also close the gap between the SCV and the shield building, but temperature effects are not considered in these analyses.) The amount and location of contact and the pressure at which it occurs all affect how the combined structure behaves. A preliminary finite element model has been developed to analyze a model of a typical steel containment vessel con-ling into contact with an outer structure. Both the steel containment vessel and the outer contact structure were modelled with axisymmetric shell finite elements. Of particular interest are the influence that the contact structure has on deformation and potential failure modes of the containment vessel. Furthermore, the coefficient of friction between the two structures was varied to study its effects on the behavior of the containment vessel and on the uplift loads transmitted to the contact structure. These analyses show that the material properties of an outer contact structure and the amount

  12. Effect of Cooling Rate on Microstructures and Mechanical Properties in SA508 Gr4N High Strength Low Alloy Steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minchul; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The microstructure of Ni-Cr-Mo low alloy steel is a mixture of tempered martensite and tempered lower bainite and that of Mn-Mo-Ni low alloy steel is predominantly tempered upper bainite. Higher strength and toughness steels are very attractive as an eligible RPV steel, so several researchers have studied to use the Ni-Cr-Mo low alloy steel for the NPP application. Because of the thickness of reactor vessel, there are large differences in austenitizing cooling rates between the surface and the center locations of thickness in RPV. Because the cooling rates after austenitization determine the microstructure, it would affect the mechanical properties in Ni-Cr-Mo low alloy steel, and it may lead to inhomogeneous characteristics when the commercial scale of RPV is fabricated. In order to apply the Ni-Cr-Mo low alloy steel to RPV, it is necessary to evaluate the changes of microstructure and mechanical properties with varying phase fractions in Ni-Cr-Mo low alloy steel. In this study, the effects of martensite and bainite fractions on mechanical properties in Ni-Cr-Mo low alloy steel were examined by controlling the cooling rate after austenitization. First of all, continuous cooling transformation(CCT) diagram was established from the dilatometric analyses. Then, the phase fractions at each cooling rate were quantitatively evaluated. Finally, the mechanical properties were correlated with the phase fraction, especially fraction of martensite in Ni-Cr-Mo low alloy steel.

  13. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    International Nuclear Information System (INIS)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-01-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels. (orig.)

  14. VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation)], E-mail: bechta@sbor.spb.su; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almiashev, V.I. [Institute of Silicate Chemistry, Russian Academy of Sciences (ISCh RAS), St. Petersburg (Russian Federation); Lopukh, D.B. [SPb State Electrotechnical University (SPbGETU), St. Petersburg (Russian Federation); Bottomley, D. [EUROPAISCHE KOMMISSION, Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA/DEN/DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Rossendorf (FZR), Dresden (Germany); Fichot, F. [IRSN/DPAM/SEMCA, St. Paul lez Durance (France); Kymalainen, O. [FORTUM Nuclear Services Ltd., Espoo (Finland)

    2009-06-15

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  15. VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

    International Nuclear Information System (INIS)

    Bechta, S.V.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

    2009-01-01

    The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

  16. Applicability of newly developed 610MPa class heavy thickness high strength steel to boiler pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Katayama, Norihiko; Kaihara, Shoichiro; Ishii, Jun [Ishikawajima-Harima Heavy Industries Corp., Yokohama (Japan); Kajigaya, Ichiro [Ishikawajima-Harima Heavy Industries Corp., Tokyo (Japan); Totsuka, Takehiro; Miyazaki, Takashi [Ishikawajima-Harima Heavy Industries Corp., Aioi (Japan)

    1995-11-01

    Construction of a 350 MW Class PFBC (Pressurized Fluidized Bed Combustion) boiler plant is under planning in Japan. Design temperature and pressure of the vessel are maximum 350 C and 1.69 MPa, respectively. As the plate thickness of the vessel exceeds over 100 mm, high strength steel plate of good weldability and less susceptible to reheat cracking was required and developed. The steel was aimed to satisfy the tensile strength over 610 MPa at 350 C after postweld heat treatment (PWHT), with good notch toughness. The authors investigated the welding performances of the newly developed steel by using 150 mm-thick plate welded by pulsed-MAG and SAW methods. It was confirmed that the newly developed steel and its welds possess sufficient strength and toughness after PWHT, and applicable to the actual pressure vessel.

  17. Development of improved SGV480 steel plate for containment vessel in PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, Norioki [Advanced Nuclear Equipment Research Inst., Tokyo (Japan); Morikage, Yasushi; Okayama, Yutaka; Higashikubo, Tomohiro

    2001-01-01

    When a nuclear containment vessel made of steel plate at PWR plants in Japan is produced, SGV480 steel plate made by annealing method according to JIS G3118 is usually used in main. And, when thickness of welding portion of the vessel is larger than 38 mm, as heat treatment after welding is regulated to carry out according to the ministerial ordinance, it is difficult in actual to carry out the heat treatment of the actual welded portions. In a leading plant, approval of welding using a special method without heat treatment less than 47.25 mm of SGV480 carbon steel plate for JIS G3118 middle and ordinary pressure vessel was carried out to supply it for actual use. And, it is required for protection of welding fracture to carry out pre-heat treatment before welding. Because of increasing plate thickness requiring for lower temperature and more seismic resistance in construction condition, in order to produce a containment vessel without heat treatment after welding, more toughness is required for using material and welded portion. Therefore, a new SGV480 steel plate was developed by using TMCP method of modern steel manufacturing technology, to establish lower carbon equivalence and finer texture with upgrading of both toughness and weldability, without heat treatment after welding and pre-heat treatment before welding, at the Shin-Nippon Steel Co, Ltd. and Kawasaki Steel, Co. Ltd., respectively. (G.K.)

  18. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1993-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (50--90mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments

  19. Neutron irradiation effects on mechanical properties in SA508 Gr4N high strength low alloy steel

    International Nuclear Information System (INIS)

    Kim, Minchul; Lee, Kihyoung; Park, Sanggyu; Choi, Kwonjae; Lee, Bongsang

    2012-01-01

    The Reactor Pressure Vessel (RPV) is the key component in determining the lifetime of nuclear power plants because it is subject to the significant aging degradation by irradiation and thermal aging, and there is no practical method for replacing that component. Advanced reactors with much larger capacity than current reactor require the usage of higher strength materials inevitably. The SA508 Gr.4N Ni Cr Mo low alloy steel, in which Ni and Cr contents are larger than in conventional RPV steels, could be a promising RPV material offering improved strength and toughness from its tempered martensitic microstructure. For a structural integrity of RPV, the effect of neutron irradiation on the material property is one of the key issues. The RPV materials suffer from the significant degradation of transition properties by the irradiation embrittlement when its strength is increased by a hardening mechanism. Therefore, the potential for application of SA508 Gr.4N steel as the structural components for nuclear power reactors depends on its ability to maintain adequate transition properties against the operating neutron does. However, it is not easy to fine the data on the irradiation effect on the mechanical properties of SA508 Gr.4N steel. In this study, the irradiation embrittlement of SA508 Gr.4N Ni Cr Mo low alloy steel was evaluated by using specimens irradiated in research reactor. For comparison, the variations of mechanical properties by neutron irradiation for commercial SA508 Gr.3 Mn Mo Ni low alloy steel were also evaluated

  20. Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

    For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

  1. Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

    International Nuclear Information System (INIS)

    Krasikov, E. A.

    2012-01-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated

  2. Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

    International Nuclear Information System (INIS)

    Krasikov, E.A.

    2012-01-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 o C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 o C and following extra irradiation (87 h at 330 o C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which, together with associated management methods, will help

  3. Controlling RPV embrittlement through wet annealing in support of life attainment and life extension decisions

    Energy Technology Data Exchange (ETDEWEB)

    Krasikov, E. A. [National Research Centre Kurchatov Inst., 1, Kurchatov Sq., Moscow, 123182 (Russian Federation)

    2012-07-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of Nuclear Power Plant (NPP) safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. Low temperature 'wet' annealing at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. As a rule there is no recovery effect up to annealing and irradiation temperature difference of 70 deg. C. It is known, however, that along with radiation embrittlement neutron irradiation may mitigate the radiation damage in metals. Therefore we have tried to test the possibility to use the effect of radiation-induced ductilization in 'wet' annealing technology by means of nuclear heat utilization as heat and neutron irradiation sources at once. In support of the above-mentioned conception the 3-year duration reactor experiment on 15Cr3NiMoV type steel with preliminary irradiation at operating Pressurized Water Reactor (PWR) at 270 deg. C and following extra irradiation (87 h at 330 deg. C) at IR-8 test reactor was fulfilled. In fact, embrittlement was partly suppressed up to value equivalent to 1,5 fold neutron fluence decrease. The degree of recovery in case of radiation enhanced annealing is equal to 27% whereas furnace annealing results in zero effect under existing conditions. Mechanism of the radiation-induced damage mitigation is proposed. It is hoped that 'wet' annealing technology will help provide a better management of the RPV degradation as a factor affecting the lifetime of nuclear power plants which

  4. 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

  5. Hydrogen induced plastic damage in pressure vessel steel of 2.25Cr-1Mo

    International Nuclear Information System (INIS)

    Han, G.W.; Song, Y.J.

    1995-01-01

    2.25Cr-1Mo steel is generally employed as a hydrogenation reaction vessel material used at elevated temperature and in a hydrogen containing environment. During service of the reaction vessel, a large number of hydrogen atoms would enter its wall. When the reaction vessel is shutdown and the temperature reduces to about ambient temperature, the hydrogen atoms remaining in the wall would induce plastic damage in the steel. The mechanism of hydrogen induced plastic damage is different for various materials with different microstructures. Investigations have demonstrated that the hydrogen induced plastic damage in carbide annealed carbon steels is caused by hydrogen accelerating the initiating and growing of microvoids from the carbide particles. However, SEM examination on the fracture surface of hydrogen charged tensile specimen of 2.25Cr-1Mo steel show that a large number of fisheyes appear on the fracture surface. This indicates that hydrogen induced plastic damage in 2.25Cr-1Mo steel is related to the occurrence of fisheye cracks during plastic deformation. By means of micro-fracture mechanics to analyze fisheye crack occurrence from the first generation microvoid, the mechanism of hydrogen induced plastic damage in the pressure vessel steel is investigated

  6. Product consistency testing of three reference glasses in stainless steel and perfluoroalkoxy resin vessels

    International Nuclear Information System (INIS)

    Olson, K.M.; Smith, G.L.; Marschman, S.C.

    1995-03-01

    Because of their chemical durability, silicate glasses have been proposed and researched since the mid-1950s as a medium for incorporating high-level radioactive waste (HLW) generated from processing of nuclear materials. A number of different waste forms were evaluated and ranked in the early 1980s; durability (leach resistance) was the highest weighted factor. Borosilicate glass was rated the best waste form available for incorporation of HLW. Four different types of vessels and three different glasses were used to study the possible effect of vessel composition on durability test results from the Production Consistency Test (PCT). The vessels were 45-m 304 stainless steel vessels, 150-m 304 L stainless steel vessels, and 60-m perfluoroalkoxy (PFA) fluoropolymer resin vessels. The three glasses were the Environmental Assessment glass manufactured by Corning Incorporated and supplied by Westinghouse Savannah River company, and West Valley Nuclear Services reference glasses 5 and 6, manufactured and supplied by Catholic University of America. Within experimental error, no differences were found in durability test results using the 3 different glasses in the 304L stainless steel or PFA fluoropolymer resin vessels over the seven-day test period

  7. Radiation embrittlement of WWER-1000 reactor vessel steels

    International Nuclear Information System (INIS)

    Nikolaeva, A.V.; Nikolaev, Yu.A.; Kevorkyan, Yu.R.

    2001-01-01

    Results obtained on the blank samples of materials of the WWER-1000 vessels irradiated by low density neutron flux are discussed. Chemical composition of the materials is characterized by the low content of the impurities (copper and phosphorus) and high content of nickel. Dependence of the radiation embrittlement of the WWER-1000 vessel materials on metallurgic variables and damage dose is treated. The research showed that nickel largely enhanced the radiation embrittlement. New dependences for determination of the radiation embrittlement real rate of the WWER-1000 vessel materials and its conservative estimation were developed [ru

  8. Joining dissimilar stainless steels for pressure vessel components

    International Nuclear Information System (INIS)

    Zheng Sun; Huai-Yue Han

    1994-01-01

    A series of studies was carried out to examine the weldability and properties of dissimilar steel joints between martensitic and austenitic stainless steels - F6NM (OCrl3Ni4Mo) and AISI 347, respectively. The weldability tests included weld thermal simulation of the martensitic steel for investigating the influence of weld thermal cycles and post-weld heat treatment (PWHT) on the mechanical properties of the heat-affected zone (HAZ); implant testing for examining the tendency for cold cracking of martensitic steel; rigid restraint testing for determining hot crack susceptibility of the multi-pass dissimilar steel joints. The joints were subjected to various mechanical tests including a tensile test, bending test and impact test at various temperatures, as well as slow strain-rate test for examining the stress corrosion cracking tendency in the simulated environment of a primary circuit of a PWR. Based on the weldability tests, a welding procedure - tungsten inert gas (TIG) welding for root passes with HNiCrMo-2B wire followed by manual metal arc (MMA) welding using coated electrode ENiCrFe-3B - was developed and a PWHT at 600 deg C/2h was recommended. Furthermore, the welding of tube/tube joints between these dissimilar steels is described. (21 refs., 11 figs., 14 tabs.)

  9. Study of radiation damage of steels for light water pressure vessels at UJV

    International Nuclear Information System (INIS)

    Vacek, N.; Stoces, B.

    1980-01-01

    Preoperational determination of radiation resistance of pressure vessel steels is performed at accelerated neutron exposure in a test or materials research reactor. The results obtained at accelerated and operating exposure are not fully identical and surveillance bodies are therefore used manufactured from the pressure vessel material. Currently, the following steels are used for the manufacture of light water reactor pressure vessels: Mn-Mo-Ni (ASTM-A533-B, ASTM-A508), Cr-Mo-V (15Kh2M1FA). At UJV Rez, for irradiation Chanca-M probes imported from France are used featuring electric temperature control. Almost identical radiation embrittlement was measured for all three steels after irradiation with a neutron fluence of 3x10 23 n.m -2 at a temperature of 290 degC. (H.S.)

  10. R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F.

    2000-01-01

    Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study thye evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (Author)

  11. Characterization of the weld HAZ properties of nuclear reactor pressure vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joo Hag; Shin, H. S.; Moon, J. G. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    This work contains an investigation on the microstructure and toughness in the weld heat-affected zone (HAZ) of a quenched and tempered SA 508 Cl. 3 reactor pressure vessel (RPV) steel. In order to evaluate systematically the notch toughness and microstructural alterations, a unit HAZ concept was applied to the multipass weld HAZ of RPV steel. Seven typical positions were selected to evaluate the spatial distribution of notch toughness and microstructure in the unit HAZ. As a result of notch toughness evaluation, three coarse-grained regions and two fine-grained regions of SA 508 Cl. 3 RPV steel HAZ showed relatively good toughness. On the contrary, an intercritically reheated and a subcritically reheated region showed lower toughness than the base metal. The region which first and second peak temperatures are 700 deg C showed the lowest toughness among the low toughness region because of carbide coarsening. Therefore, it was proposed that the notch position in the surveillance HAZ specimen should be placed to the boundary between the HAZ and the base metal. The method, which evaluates the fracture toughness in the transition region of ferritic steel, was effectively applicable to the various HAZ regions of RPV steel. The fracture toughness test results were nearly same as the notch toughness test results. The volume fraction of tempered martensite phase was revealed as the most dominant factor that determines fracture toughness. 59 refs., 29 figs., 10 tabs. (Author)

  12. Heissdampfreaktor (HDR) steel-containment-vessel and floodwater-storage-tank structural-dynamics tests

    International Nuclear Information System (INIS)

    Arendts, J.G.

    1982-01-01

    Inertance (vibration) testing of two significant vessels at the Heissdampfreaktor (HDR) facility, located near Kahl, West Germany, was recently completed. Transfer functions were obtained for determination of the modal properties (frequencies, mode shapes and damping) of the vessels using two different test methods for comparative purposes. One of the vessels tested was the steel containment vessel (SCV). The SCV is approximately 180 feet high and 65 feet in diameter with a 1.2-inch wall thickness. The other vessel, called the floodwater storage tank (FWST), is a vertically standing vessel approximately 40 feet high and 10 feet in diameter with a 1/2-inch wall thickness. The FWST support skirt is square (in plan views) with its corners intersecting the ellipsoidal bottom head near the knuckle region

  13. Effects of Microstructural Inhomogeneity on Charpy Impact Properties for Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Song, Jaemin; Kim, Min-Chul; Choi, Kwon-Jae; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Reactor pressure vessel (RPV) steels are fabricated by vacuum carbon deoxidation (VCD), and then heat treatment of quenching and tempering is conducted after forging. The through-the-thickness variation of microstructure in RPV can occur due to the cooling rate gradient during quenching and inhomogeneous deformation during forging process. The variation of microstructure in RPV affects the mechanical properties, and inhomogeneity in mechanical properties can occur. The evaluation of mechanical properties of RPV is conducted at thickness of 1/4T. In order to evaluate the safety of RPV more correctly, the research about the through-the-thickness variation of microstructure and mechanical properties in RPV is need. 1. The fine low bainite (LB) is the dominant phase at the inner-surface (0T), but coarse upper bainite (UB) is the dominant phase at the center (1/2T). This is because cooling rate gradient from surface to center occurs during quenching. 2. Inter-lath carbides act as fracture initiation site, and it reduces impact toughness. 3. The upper shelf energy is low and the reference temperatures are high at the 1/4T. Impact properties are poor at 1/4T because of the formation of coarse upper bainite structure and coarse inter-lath carbides.

  14. The influence of the crust layer on RPV structural failure under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Mao, Jianfeng, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Li, Xiangqing [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Bao, Shiyi [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, Lijia [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Gao, Zengliang [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2017-05-15

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  15. The influence of the crust layer on RPV structural failure under severe accident condition

    International Nuclear Information System (INIS)

    Mao, Jianfeng; Li, Xiangqing; Bao, Shiyi; Luo, Lijia; Gao, Zengliang

    2017-01-01

    Highlights: • The crust layer greatly affects the RPV structural behavior. • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage mainly contribute to RPV failure. • An elastic core in RPV wall is essential for ensuring RPV integrity. • The multiaxial state of stress accelerates the total damage evolution. - Abstract: The so called ‘in-vessel retention (IVR)’ is regarded as a severe accident (SA) mitigation strategy, which is widely used in most of advanced nuclear power plants. The effectiveness of IVR strategy is to employ the external water flooding to cool the reactor pressure vessel (RPV). The RPV integrity has to be maintained within a required period during the IVR period. The degraded melting core is assumed to be arrested in the lower head (LH) to form the melting pool that is bounded by upper, side and lower crusts. Consequently, the existence of the crust layer greatly affects the RPV structural behavior as well as failure process. In order to disclose this influence caused by the crust layer, a detailed investigation is conducted by using numerical simulation on the two RPVs with and without crust layer respectively. Taking the RPV without crust layer as a basis for the comparison, the present study assesses the likelihood and potential failure location, time and mode of the LH under the loadings of the critical heat flux (CHF) and slight internal pressure. Due to the high temperature melt on the inside and nucleate boiling on the outside, the RPV integrity is found to be compromised by melt-through, creep, elasticity, plasticity as well as thermal expansion. Through in-depth investigation, it is found that the creep and plasticity are of vital importance to the final structural failure, and the introduction of crust layer results in a significant change on field parameters in terms of temperature, deformation, stress(strain), triaxiality factor and total damage.

  16. Effect of stress relief parameters on the mechanical properties of pressure vessel steels and weldments

    International Nuclear Information System (INIS)

    Canonico, D.A.; Stelzman, W.J.

    1976-01-01

    Post weld heat treatments of thick-section A533B steel for nuclear pressure vessels are discussed with reference to the ASME code. The discussion is in the form of a lecture and summarized by noting that the ASME code, in particular Section III, Division 1, imposes a post weld heat treatment requirement on pressure vessels fabricated from low alloy high strength steels. The Code permits a holding temperature range, the high side of which could result in poorer toughness properties. Long times in excess of 100 hours and/or high temperatures, 649 0 C can result in an increase in the NDT and a decrease in the upper shelf energy

  17. Applicability of JIS SPV 50 steel to primary containment vessels of nuclear power stations

    International Nuclear Information System (INIS)

    Iida, K.; Ishikawa, K.; Satoh, M.; Soya, I.

    1980-01-01

    The fracture toughness of JIS SPV 50 steel and its weldment has been examined in order to verify the applicability of these materials to primary containment vessels of nuclear power stations. Test results were evaluated using elastic plastic fracture mechanics through the COD and the J integral concepts for non ductile fracture initiation characteristics. Linear fracture mechanics was employed for propagation arrest characteristics. Results showed that the materials tested here have a sufficient fracture toughness to prevent nonductile fracture and that this steel is a suitable material for use in construction of primary containment vessels of nuclear power stations. (author)

  18. Neutron flux effect on the fracture toughness behavior of Tihange-III RPV material

    International Nuclear Information System (INIS)

    Gerard, R.; Chaouadi, R.; Bertolis, D.

    2015-01-01

    The question whether material test reactor (MTR) data can be used to supplement power reactor pressure vessel (RPV) surveillance data is still debated in the international community and its implications are particularly important in the perspective of long term operation (LTO). However, addressing the flux effect can be confusing if specific material and irradiation variables are not taken into account. This means that the answer to whether there is flux effect or not is neither 'no' nor 'yes' without specifying the application range. Indeed, neutron flux effect was recognized to occur in high Cu-containing steels in the low fluence range. But at high fluence, relevant for long term operation, it becomes difficult to clearly distinguish the differences between high flux and low flux. In this work, we irradiated the low Cu base metal and weld of the Tihange-III surveillance coupon in the BR2 reactor at high flux. The BR2 flux is about two orders of magnitude higher than the flux in the surveillance position. Tensile, Charpy impact and fracture toughness tests were performed on both the surveillance and MTR specimens and compared to assess the neutron flux effect. The results confirm that, at high fluence levels, the flux effect on mechanical properties is not significant, offering therefore the possibility of accelerated irradiation to investigate RPV embrittlement in the high fluence regime relevant for long term operation. (authors)

  19. Effect of solute elements on hardening of thermally-aged RPV model alloys

    International Nuclear Information System (INIS)

    Nomoto, A.; Nishida, K.; Dohi, K.; Soneda, N.; Liu, L.; Sekimura, N.; Li, Z.

    2015-01-01

    Embrittlement correlation methods for irradiated reactor pressure vessel (RPV) steels have been developed worldwide to predict the amount of embrittlement during plant operation. The effect of chemical composition on embrittlement is not fully understood, particularly the process of solute atom behavior during solute atom formation. In this series of slides we report the results of thermal ageing experiments of RPV model alloys in order to obtain information on the effect of chemical composition on the hardening process. We can draw the following conclusions. First, the addition of Ni or Si alone to Fe-Cu model alloys does not have clear effect but the addition of Mn to Fe-Cu-Ni alloy accelerates the cluster formation and hardening drastically, the effect of composition on the cluster strength is not clear. Secondly, the hardening process before the hardening peak has linear correlation with APT (Atom Probe Tomography) results and the RSS (Root-Sum-Square)sum model seems to explain the relationship between increase in hardness and APT data in a more consistent manner

  20. Experimental Study of Interactions Between Sub-oxidized Corium and Reactor Vessel Steel

    International Nuclear Information System (INIS)

    Bechta, S.V.; Khabensky, V.B.; Granovsky, V.S.; Krushinov, E.V.; Vitol, S.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Fischer, M.; Piluso, P.; Altstadt, E.; Willschutz, H.G.; Fichoti, F.

    2006-01-01

    One of the critical factors in the analysis of in-vessel melt retention is the vessel strength. It is, in particular, sensitive to the thickness of intact vessel wall, which, in its turn, depends on the thermal conditions and physicochemical interactions with corium. Physicochemical interaction of prototypic UO 2 -ZrO 2 -Zr corium melt and VVER vessel steel was examined during the 2. Phase of the ISTC METCOR Project. Rasplav-3 test facility was used for conducting four tests, in which the Zr oxidation degree and interaction front temperature were varied; in one of the tests, stainless steel was added to the melt. Direct experimental measurements and post-test analyses were used for determining corrosion kinetics and maximum corrosion depth (i.e. the physicochemical impact of corium on the cooled vessel steel specimens), as well as the steel temperature conditions during the interaction, and finally the structure and composition of crystallized ingots, including the interaction zone. The minimum temperature on the interaction front boundary, which determined its final position and maximum corrosion depth was ∼ 1090 deg. C. An empirical correlation for calculation of corrosion kinetics has been derived. (authors)

  1. Ultimate limit states of steel containment vessel under earthquake loadings

    International Nuclear Information System (INIS)

    Akiyama, Hiroshi; Yuhara, Tetsuo; Shimizu, Seiichi; Hayashi, Kazutoshi; Takahashi, Tadao.

    1986-01-01

    The limit state induced by buckling of cylindrical steel structures under earthquake loadings was investigated from the standpoint of energy concept. A number of the buckling test of cylindrical steel shell structures has been made, which showed that they have a stable load-displacement relation and adequate deformation capacities beyond the buckling. The authors are proposing that the energy input imparted by strong earthquakes to buckled structures and the deformation capacity in post-buckling are suitable indices for seismic resistance of the cylindrical steel shell structures because the buckling does not cause the structure immediately to collapse in the case of such repeated loading as earthquake motions. The purpose of this study is to investigate the energy input to buckled cylindrical steel structures with an increase in the intensity of earthquake motions. A series of nonlinear dynamic analyses were performed under various types of earthquake records by using a hysteresis loop, including buckling, which was derived from the buckling tests. The limit state could be defined as the state in which the deformation of and the energy input into buckled structures increase divergently when the intensity of the earthquake excitation exceeds a certain value. The results obtained in this paper are intended to be adopted to the limit state in the post-buckling region to evaluate the margin of safety against the buckling resistance of cylindrical steel structures under strong earthquake loadings. (author)

  2. Characterization of four prestressed concrete reactor vessel liner steels

    International Nuclear Information System (INIS)

    Nanstad, R.K.

    1980-12-01

    A program of fracture toughness testing and analysis is being performed with PCRV steels for HTGRs. This report focuses on background information for the base materials and results of characterization testing, such as tensile and impact properties, chemical composition, and microstructural examination. The steels tested were an SA-508 class 1 forging, two plates of SA-537 class 1, and one plate of SA-537 class 2. Tensile requirements in effect at the time of procurement are met by all four steels. However, the SA-537 class 2 plate would not meet the minimum requirement for yield strength. Drop-weight and Charpy impact tests verified that the RT/sub NDT/ is equal to the NDT for each steel. Charpy impact energies at the NDT range from 40 J (30 ft-lb) for one heat of SA-537 class 1 to 100 J (74 ft-lb) for the SA-537 class 2 plate; upper-shelf energies range from 170 to 310 J (125 to 228 ft-lb) for the same two steels, respectively. The onset of upper-shelf energy occurred at temperatures ranging from 0 to 50 0 C

  3. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Kocik, J.; Keilova, E.

    1993-01-01

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs

  4. Radiation damage structure in irradiated and annealed 440 WWER-Type reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kocik, J; Keilova, E [Czech Nuclear Society, Prague (Czech Republic)

    1994-12-31

    A review of irradiation damages in WWER-type RPV steels based on conventional Transmission Electron Microscopy investigations in a power reactor and a research reactor, is presented; the samples consist in Cr-Mo-V ferritic steel (15Kh2MFA type). The visible part of radiation-induced defects consists of very fine vanadium carbide precipitates, small dislocation loops and black dots (presumably corresponding) to clusters and particle embryos formed from vacancies and solute-atoms (vanadium, copper, phosphorus) and carbon associated with vanadium. Radiation-induced defects are concentrated at dislocation substructure during irradiation in a power reactor, revealing the role of radiation-enhanced diffusion in damage structure forming process. Contrarily, the distribution of defects resulting from annealing of specimens irradiated in the research reactor is pre-determined by an homogenous distribution of radiation-induced defects prior to annealing. Increasing the number of re-irradiation and annealing cycles, the amount of dislocation loops among all defects seems to be growing. Simultaneously, the dislocation substructure recovers considerably. (authors). 14 refs., 11 figs., 3 tabs.

  5. Fracture toughness evaluation in the transition region of reactor pressure vessel steel

    International Nuclear Information System (INIS)

    Onizawa, K.; Suzuki, M.

    1995-01-01

    The fracture toughness (K jc and Jc) values at the cleavage fracture initiation in the transition region of a RPV steel were investigated using mainly precracked Charpy specimens. A conventional statistical approach and a fractographic study were applied to analyze the scatter of the fracture toughness values from precracked Charpy specimens. The material used was an ASTM A533B class 1 steel, which was designated as an IAEA correlation monitor material, JRQ. A lower bound transition curve of the fracture toughness for unirradiated condition was determined by the 5% confidence limit from the Weibull and fractographic analyses. The lower bound transition curve after irradiation was evaluated based on the statistics of unirradiated specimens. The results indicated that the shift of the fracture toughness transition curbe were somewhat larger than the Charpy 41J transition temperature. The parameters to determine the lower bound toughness such as the Weibull slope and the amount of ductile crack growth are discussed. The results are also compared with a model based on weakest link theory. (author). 12 refs, 12 figs, 5 tabs

  6. 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

  7. Spanish RPV head penetrations. Regulatory status

    International Nuclear Information System (INIS)

    Figueras, J.M.; Colino, J.R.

    1995-01-01

    The paper presents the actual status of inspection results on the Spanish PWR RPV CRD head penetrations (CRDH's), after two years of a whole program of inspections in all affected plants. Actual situation of penetrations pertaining to ALMARAZ 1 and 2, ASCO 1 and 2 and VANDELLOS 2 NPP's show any damage in those CRDH's inspected in 1993 and 1994 (roughly 20 out of 65 CRDH's at each unit). The paper presents a summary of CRDH characteristics, inspection methods and results obtained in each plant. TRILLO NPP has a different CRDH design (KWU-SIEMENS type) and for that reason is not considered an affected plant nor has conducted any inspection up to now. JOSE CABRERA (ZORITA) NPP has shown extensive damage, both in the lower side (weldment to the vessel) and in the upper free span area, near bimetallic weldment to SS 304, in active and nonactive penetrations and also in the vent nozzle. The paper comments extensively on the CRDH materials general data, root-cause analysis and structural analysis of degraded zones, inspection results, repair actions and other additional actions applied up to now. Finally, the paper deals with the regulatory actions taken by CSN on this topic, both for those NPP's actually non affected by the IGSCC phenomenon in the RPV CRDH's and for the specific safety case of ZORITA NPP. (author)

  8. Acoustic emission test on a 25mm thick mild steel pressure vessel with inserted defects

    International Nuclear Information System (INIS)

    Bentley, P.G.; Dawson, D.G.; Hanley, D.J.; Kirby, N.

    1976-12-01

    Acoustic emission measurements have been taken on an experimental mild steel vessel with 4 inserted defects ranging in severity up to 90% of through thickness. The vessel was subjected to a series of pressure excursions of increasing magnitude until failure occurred by extension of the largest inserted defect through the vessel wall. No acoustic emission was detected throughout any part of the tests which would indicate the presence of such serious defects or of impending failure. Measurements of acoustic emission from metallurgical specimens are included and the results of post test inspection using conventional NDT and metallographic techniques are reported. (author)

  9. Pressure vessels for reactors made from structural steel with limited tensile strength

    International Nuclear Information System (INIS)

    Machatti, H.

    1973-01-01

    The reactor pressure vessel is prestressed in several directions with prestressing elements fabricated of steel with a high yielding point. This design allows a substantial reduction of wall thickness or an increase of the inner diameter at equal wall thickness. The prestress of the prestressing elements is designed to achieve a maximum stress release of the vessel walls at normal operating conditions and to fully utilize the maximum load of the vessel walls. For safety reasons the cross section of the prestressing elements is constructed in a way that strain is always 20 % lower the yield point. (P.K.)

  10. Influence of heat treatments on thermoelectric power of pressure vessel steels: effect of microstructural evolutions of strongly segregated areas

    International Nuclear Information System (INIS)

    Houze, M.

    2002-12-01

    Thermoelectric power measurement (TEP) is a very potential non destructive evaluation method considered to follow ageing under neutron irradiation of pressure vessel steel of nuclear reactor. Prior to these problems, the aim of this study is to establish correlations between TEP variations and microstructural evolutions of pressure vessel steels during heat treatments. Different steels, permitting to simulate heterogeneities of pressure vessel steels and to deconvoluate main metallurgical phenomenons effects were studied. This work allowed to emphasize effect on TEP of: austenitizing and cooling conditions and therefore of microstructure, metallurgical transformations during tempering (recovery, precipitation of alloying elements), and particularly molybdenum precipitation associated to secondary hardening, residual austenite amount or partial austenitizing. (author)

  11. Study of cladding toughness in a pressure vessel steel water reactor

    International Nuclear Information System (INIS)

    Soulat, P.; Al Mundheri, M.

    1984-12-01

    Toughness of cladding and pressure vessel steel were determined at different temperatures in order to appreciate the participation of cladding resistance against crack propagation. The toughness of cladding is comparable with typical results on austenitic welds. The test on covered CT specimens shows the possibility of having a relatively good prevision of the behaviour of a coated structure

  12. Positron annihilation and Moessbauer studies of neutron irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1990-11-01

    Positron annihilation (lifetime, Doppler broadening) and Moessbauer studies on unirradiated, neutron irradiated and neutron irradiated plus annealed reactor pressure vessel steels (Soviet type 15Kh2NMFA) are presented. The role of microstructural properties and the formation of irradiation-induced precipitates is discussed. (orig.) [de

  13. Influence of steel-making process and heat-treatment temperature on the fatigue and fracture properties of pressure vessel steels

    International Nuclear Information System (INIS)

    Koh, S. K.; Na, E. G.; Baek, T. H.; Won, S. Y.; Park, S. J.; Lee, S. W.

    2001-01-01

    In this paper, high strength pressure vessel steels having the same chemical compositions were manufactured by the two different steel-making processes, such as Vacuum Degassing(VD) and Electro-Slag Remelting(ESR) methods. After the steel-making process, they were normalized at 955 deg. C, quenched at 843 .deg. C, and finally tempered at 550 .deg. C or 450 deg. C, resulting in tempered martensitic microstructures with different yielding strengths depending on the tempering conditions. Low-Cycle Fatigue(LCF) tests, Fatigue Crack Growth Rate(FCGR) tests, and fracture toughness tests were performed to investigate the fatigue and fracture behaviors of the pressure vessel steels. In contrast to very similar monotonic, LCF, and FCGR behaviors between VD and ESR steels, a quite difference was noticed in the fracture toughness. Fracture toughness of ESR steel was higher than that of VD steel, being attributed to the removal of impurities in steel-making process

  14. Heavy-section steel irradiation program. Progress report, October 1992--March 1993

    International Nuclear Information System (INIS)

    Corwin, W.R.

    1998-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is one of only two more safety-related components of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established at Oak Ridge National Laboratory (ORNL) under sponsorship of the Nuclear Regulatory Commission (NRC). The primary goal of this major safety program is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior (in particular, the fracture toughness properties) of typical pressure-vessel steels as they relate to light-water-reactor pressure-vessel integrity. The program centers on experimental assessments of irradiation-induced embrittlement (including the completion of certain irradiation studies previously conducted by the Heavy-Section Steel Technology Program) augmented by detailed examinations and modeling of the accompanying microstructural changes. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties

  15. Large inelastic deformation analysis of steel pressure vessels at high temperature

    International Nuclear Information System (INIS)

    Ikonen, K.

    2001-01-01

    This publication describes the calculation methodology developed for a large inelastic deformation analysis of pressure vessels at high temperature. Continuum mechanical formulation related to a large deformation analysis is presented. Application of the constitutive equations is simplified when the evolution of stress and deformation state of an infinitesimal material element is considered in the directions of principal strains determined by the deformation during a finite time increment. A quantitative modelling of time dependent inelastic deformation is applied for reactor pressure vessel steels. Experimental data of uniaxial tensile, relaxation and creep tests performed at different laboratories for reactor pressure vessel steels are investigated and processed. An inelastic deformation rate model of strain hardening type is adopted. The model simulates well the axial tensile, relaxation and creep tests from room temperature to high temperature with only a few fitting parameters. The measurement data refined for the inelastic deformation rate model show useful information about inelastic deformation phenomena of reactor pressure vessel steels over a wide temperature range. The methodology and calculation process are validated by comparing the calculated results with measurements from experiments on small scale pressure vessels. A reasonably good agreement, when taking several uncertainties into account, is obtained between the measured and calculated results concerning deformation rate and failure location. (orig.)

  16. Applicability of JIS SPV 50 steel to primary containment vessel of nuclear power station

    International Nuclear Information System (INIS)

    Iida, Kunihiro; Ishikawa, Koji; Sakai, Keiichi; Onozuka, Masakazu; Sato, Makoto.

    1979-01-01

    The space within reactor containment vessels must be expanded in order to improve the reliability of nuclear power plants, accordingly the adoption of large reactor containment vessels is investigated. SGV 42 and 49 steels in JIS G 3118 have been used for containment vessels so far, but stress relief annealing is required when the thickness exceeds 38 mm. The time has come when the use of thicker conventional plates without stress relieving or the use of high strength steel must be examined in detail. In this study, the tests of confirming material properties were carried out on SPV 50 in JIS G 3115, Steels for pressure vessels, aiming at the method of fabrication without stress relieving. The highest and lowest temperatures in use were set at 171 deg and -8 deg C, respectively. The chemical composition and the mechanical properties of the plates tested, the method of welding, the results of tensile test on the parent metal and the welds, the required lowest preheating temperature, the fracture toughness at low temperature and the brittle fracture causing test are reported. The parent metal and the welded joints of SPV 50 have the properties suitable to reactor containment vessels, namely the sufficient fracture toughness to guarantee the prevention of unstable fracture when the method of welding without stress relieving is adopted. (Kako, I.)

  17. 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.

  18. A three-dimensional rupture analysis of steel liners anchored to concrete pressure and containment vessels

    International Nuclear Information System (INIS)

    Bangash, Y.

    1987-01-01

    Steel liners or plates are anchored to concrete pressure and containment vessels for nuclear and offshore facilities. Due to extreme loading conditions a liner may buckle due to the pull-out or shearing of anchors from the base metal and concrete. Under certain conditions attributed to loadings, liner metal deterioration and cracking of concrete behind the liner, the liner may fail by rupture. This paper presents a three-dimensional analysis of steel-concrete elements, using finite elements analysis in which a provision is made for liner instability, anchor strength and stiffness, concrete cracking and finally liner rupture. The analysis is tested first on an octagonal slab with and without an anchored steel liner. It is then extended to concrete pressure and containment vessels. The analytical results obtained are compared well with those available from the experimental tests and other sources. (author)

  19. Some aspects of experimental investigation of the RPV material properties

    International Nuclear Information System (INIS)

    Lipka, J.; Hascik, J.; Groene, R.; Slugen, V.; Vitazek, K.; Hinca, R.; Toth, I.; Kupca, L.

    1996-01-01

    Moessbauer spectra (MS) and Electron-Positron Annihilation (EPA) spectra at room temperature have been measured on the samples from Reactor Pressure Vessel (RPV). Both types of measurements showed that the changes associated with the effects of neutron irradiation, as well as thermal treatment, can be detected by Moessbauer and Electron-Positron Annihilation spectroscopy. On base of a positive results achieved in MS and EPA measurements the complementary surveillance specimen program for the Reactor Pressure Vessel Materials Study of the third and fourth units NPP Jaslovske Bohunice has been prepared. The complementary surveillance specimen program has started in May 1995. The samples with proper design from basic and welded RPV materials were measured by MS and EPA before placing into the reactor. After neutron irradiation the samples become radioactive because of 59 Co content. To eliminate the influence of 60 Co gamma radiation on the EPA angular correlation and time spectra a three detectors spectrometer has been introduced. (author)

  20. Irradiation effects in strain aged pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M; Myers, H P

    1962-02-15

    Tensile specimens, Charpy-V notch and subsize impact specimens of an aluminium killed carbon manganese steel, have been irradiated at 160 - 190 deg C in the reactor G1. The total neutron dose received was 2.4 x 10{sup 18} n/cm{sup 2} (> 1 MeV). Specimens were prepared from normalized plate and from strain aged material from the same plate. It was found that the changes in brittle ductile transition temperature due to neutron irradiation and those due to strain ageing must be considered additive.

  1. Electrode for welding steel for WWER-1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Lakatos, L.

    Of two types of electrodes, ie., with an alloyed core and with an unalloyed core, an electrode was chosen consisting of a basic coat and an unalloyed core. Fluctuations are shown of shear strength, tensile strenght and contraction with the welding mode and annealing temperature. It was found that pre-heating to 250 and 350 degC, respectively, was most suitable for welding a pressure vessel manufactured from material designated SKODA A3/II. Annealing aimed at removing stress was chosen at 650 to 700 degC. (H.S.)

  2. 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.

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

    International Nuclear Information System (INIS)

    Chakraborty, Pritam; Biner, Suleyman Bulent; Zhang, Yongfeng; Spencer, Benjamin Whiting

    2015-01-01

    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.

  4. Ultimate Pressure Capacity of Prestressed Concrete Containment Vessels with Steel Fibers

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The ultimate pressure capacity (UPC) of the prestressed concrete containment vessel (PCCV) is very important since the PCCV are final protection to prevent the massive leakage of a radioactive contaminant caused by the severe accident of nuclear power plants (NPPs). The tensile behavior of a concrete is an important factor which influence to the UPC of PCCVs. Hence, nowadays, it is interested that the application of the steel fiber to the PCCVs since that the concrete with steel fiber shows an improved performance in the tensile behavior compared to reinforced concrete (RC). In this study, we performed the UPC analysis of PCCVs with steel fibers corresponding to the different volume ratio of fibers to verify the effectiveness of steel fibers on PCCVs

  5. Apparent embrittlement saturation and radiation mechanisms of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Pachur, D.

    1981-01-01

    The irradiation and annealing results of three different reactor pressure vessel steels are reported. Steel A, a basic material according to ASTM A-533 B having 0.15 percent vanadium; and Steel C contained 3.2 percent nickel. The steels were irradiated at 150, 300, and 400 degree C with neutron fluxes of 6 multiplied by 10 11 and 3 multiplied by 10 13 neutrons (n)/cm 2 /s. An apparent saturation-in-irradiation effect was found within certain neutron fluence ranges. During the annealing, various recovery processes occur in different temperature ranges. These are characterized by various activation energies. The individual processes were determined by the different time dependencies at various temperatures. Two causes for the apparent saturation were discovered from the behavior of the annealing curves

  6. Assessment of environmentally assisted cracking in PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Tice, D.R.

    1991-01-01

    There is a possibility that extension of pre-existing flaws in the reactor pressure vessel of a pressurised water reactor (PWR) may occur by environmentally assisted cracking, in particular by corrosion fatigue under cyclic transient loading. Crack growth predictions have usually been carried out using cyclic crack growth rate (da/dN) versus stress intensity range (δK) curves, such as those given in Section XI, Appendix A of the ASME Boiler and Pressure Vessel Code. However, the inherent time dependent nature of environmental cracking processes renders such an approach unrealistic. The present paper describes the development of an alternative time based assessment methodology. Illustrative calculations of expected crack growth of assumed defects made using the cyclic (ASME XIA) and time-based approaches are compared. The results illustrate that crack growth predicted by the time-based approach can be greater or less than that calculated by the traditional method. For a PWR operated with good control of water chemistry, actual crack growth rates are expected to be well below those predicted by the ASME code. (Author)

  7. APFIM investigation of clustering in neutron-irradiated Fe-Cu alloys and pressure vessel steels

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Blavette, D.

    1996-01-01

    Pressure vessel steels used in PWRs are known to be prone to hardening and embrittlement under neutron irradiation. The changes in mechanical properties are commonly supposed to result from the formation of point defects, dislocation loops, voids and copper-rich precipitates. However, the real nature of the irradiation induced damage, in these particularly low copper steels (>0,1 wt%), has not been clearly identify yet. A new experimental work has been carried out thanks to atom probe and field ion microscopy (APFIM) facilities and, more particularly with a new generation of atom probe recently developed, namely the tomographic atom probe (TAP), in order to improve: the understanding of the complex behavior of copper precipitation which occurs when low-alloyed Fe-Cu model alloys are irradiated with neutrons; the microstructural characterization of the pressure vessel steel of the CHOOZ A reactor under various fluences (French Surveillance Programme). The investigations clearly reveal the precipitation of copper-rich clusters in irradiated Fe-Cu alloys while more complicated Si, Ni, Mn and Cu-solute 'clouds' were observed to develop in the low-copper ferritic solid solution of the pressure vessel steel. (authors)

  8. Correlation between radiation damage and magnetic properties in reactor vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Kempf, R.A., E-mail: kempf@cnea.gov.ar [División Caracterización, GCCN, CAC-CNEA (Argentina); Sacanell, J. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Milano, J. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Guerra Méndez, N. [Departamento Física de la Materia Condensada, GIyA, CAC-CNEA, CONICET (Argentina); Winkler, E.; Butera, A. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Troiani, H. [División Física de Metales, CAB-CNEA and Instituto Balseiro (UNCU), CONICET (Argentina); Saleta, M.E. [División Resonancias Magnéticas, CAB-CNEA, CONICET (Argentina); Fortis, A.M. [Departamento Estructura y Comportamiento. Gerencia Materiales-GAEN, CAC-CNEA (Argentina)

    2014-02-01

    Since reactor pressure vessel steels are ferromagnetic, provide a convenient means to monitor changes in the mechanical properties of the material upon irradiation with high energy particles, by measuring their magnetic properties. Here, we discuss the correlation between mechanical and magnetic properties and microstructure, by studying the flux effect on the nuclear pressure vessel steel used in reactors currently under construction in Argentina. Charpy-V notched specimens of this steel were irradiated in the RA1 experimental reactor at 275 °C with two lead factors (LFs), 93 and 183. The magnetic properties were studied by means of DC magnetometry and ferromagnetic resonance. The results show that the coercive field and magnetic anisotropy spatial distribution are sensitive to the LF and can be explained by taking into account the evolution of the microstructure with this parameter. The saturation magnetization shows a dominant dependence on the accumulated damage. Consequently, the mentioned techniques are suitable to estimate the degradation of the reactor vessel steel.

  9. Assessment of environmentally assisted cracking in PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Tice, D.R.

    1987-01-01

    1) Since environmentally assisted cracking (EAC) is a time dependent process, assessment should be based on time rather than cycle dependent parameters. Thus an a/sub e/ vs a/sub i/ (or strain rate) basis for assessment should be used in preference to da/dN vs ΔK. 2) The threshold strain rate or velocity for the onset of EAC is controlled by material and environmental factors (e.g. steel sulphur content and water chemistry), and possibly by mechanical loading factors such as R ratio and load interaction effects. Above the threshold, crack growth rates are usually unacceptably rapid. 3) Sample calculations show that predicted crack growth rates using a time based model can be below or above those calculated using ASME XI depending on the value of the EAC threshold velocity but that for normal PWR operating conditions rates are likely to be below those predicted by the ASME code

  10. 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

  11. Survey of postirradiation heat treatment as a means to mitigate radiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1979-01-01

    Nuclear-radiation service typically produces a progressive reduction in the notch ductility of low-alloy steels. The reduction is manifested by a decrease in Charpy-V (Csub(v)) upper-shelf energy level and by an elevation in temperature of the ductile-to-brittle transition. Post irradiation heat treatment (annealing) is being investigated as a method for the reversal of these detrimental radiation effects for reactor-vessel steels. This study was undertaken to analyze factors which could affect annealing response, report data available to qualify suspected influences on annealing, and summarize experimental results generated for many commercially produced reactor materials and companion materials produced in the laboratory

  12. Hydrogen attack of pressure-vessel steel. Progress report, April 1, 1980-March 31, 1981

    International Nuclear Information System (INIS)

    Shewmon, P.G.

    1980-12-01

    The nucleation and growth of methane bubbles in the hydrogen attack of pressure vessel steel has been shown to obey models developed to describe the growth of bubbles limiting the creep ductility of metals. This has been done through studies of the effect of prior deformation on bubble nucleation as well as the effect of methane pressure (stress) and temperature on growth kinetics. A comprehensive model of the factors limiting growth has been developed. Its application to the hydrogen attack of a 2 1/4 Cr-1 Mo steel leads to several interesting predictions

  13. 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

  14. Evaluation of defects induced by neutron radiation in reactor pressure vessels steels

    International Nuclear Information System (INIS)

    Lopez Jimenez, J.

    1978-01-01

    We have developed a method for calculating the production of neutron induced defects (depleted zone and crowdions) in ferritic pressure vessel steels for different neutron spectra. They have been analysed both the recoil primary atoms produced by elastic and inelastic collisions with fast neutrons and the ones produced by gamma-ray emission by thermal neutron absorption. Theoretical modelling of increasing in the ductile-brittle transition temperature of ferritic steels has been correlated with experimental data at irradiation temperature up to 400 degree centigree (Author) 15 refs

  15. Relationships between Charpy impact shelf energies and upper shelf Ksub(IC) values for reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Witt, F.J.

    1983-01-01

    Charpy shelf data and lower bound estimates of Ksub(IC) shelf data for the same steels and test temperatures are given. Included are some typical reactor pressure vessel steels as well as some less tough or degraded steels. The data were evaluated with shelf estimates of Ksub(IC) up to and exceeding 550 MPa√m. It is shown that the high shelf fracture toughness representative of tough reactor pressure vessel steels may be obtained from a knowledge of the Charpy shelf energies. The toughness transition may be obtained either by testing small fracture toughness specimens or by Charpy energy indexing. (U.K.)

  16. Regulatory aspects of radiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Randall, P.N.

    1979-01-01

    One purpose of this conference, is to re-examine the conventional wisdom about neutron radiation embrittlement and the methods used to counteract embrittlement in reactor vessels. Perhaps, there have been sufficient advances in fracture mechanics, core physics, dosimetry, and physical metallurgy to permit a forward step in the quantitative treatment of the subject. Certainly this would be consistent with the position of the U.S. Nuclear Regulatory Commission (the NRC) in general. ''There has been a continued evolution toward increased specificity.'' This statement appeared in the response prepared by the staff to a request from the Commission to explain how the staff decides to apply a new requirement and to whom, i.e., to back-fit or forward-fit-only or whatever. Pressure for increased specificity, i.e., for fleshing out general design criteria, comes from ''technical surprises'' in the form of operating experiences or from research information, and from attempts to improve our confidence in the safety of plants, especially new plants. Our goal is to have anticipated and evaluated all possible modes of failure with sufficient quantitativeness that the probability of failure can be estimated with some accuracy. Failing this, regulators demand large margins of safety to cover our ignorance

  17. Special servicing equipment for reactor pressurized vessel stud hole and stud accessories

    International Nuclear Information System (INIS)

    Li Jianglian

    1999-01-01

    The author briefly introduces the design and manufacture of nuclear island special servicing equipment of Nuclear Power Institute of China. Maintenance process of reactor pressurized vessel (RPV) stud hold and stud accessories the special servicing equipment include RPV flange dummy, closed-circuit television (CCTV) inspection equipment, RPV stud hole expandable comb, RPV stud hole polisher, RPV stud hold thread lubricating equipment, RPV stud hole thread miller and RPV stud hole camera. It is presented how eight kinds of special servicing equipment perform the maintenance process concerning their function, structure, and characteristics, their practical use on site is also introduced

  18. Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

    Directory of Open Access Journals (Sweden)

    J. Mao

    2018-03-01

    Full Text Available The strategy denoted as in-vessel retention (IVR is widely used in severe accident (SA management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.

  19. Magnetic Barkhausen noise and magneto acoustic emission in pressure vessel steel

    International Nuclear Information System (INIS)

    Neyra Astudillo, Miriam Rocío; López Pumarega, María Isabel; Núñez, Nicolás Marcelo; Pochettino, Alberto; Ruzzante, José

    2017-01-01

    Magnetic Barkhausen Noise (MBN) and Magneto Acoustic Emission (MAE) were studied in A508 Class II forged steel used for pressure vessels in nuclear power stations. The magnetic experimental determinations were completed with a macro graphic study of sulfides and the texture analysis of the material. The analysis of these results allows us to determine connections between the magnetic anisotropy, texture and microstructure of the material. Results clearly suggest that the plastic flow direction is different from the forging direction indicated by the material supplier - Highlights: • MBN and MAE studied in nuclear power pressure vessel steel. • Comparison with macro graphic study of sulfides and texture analysis of the material. • Connections with magnetic anisotropy, texture and microstructure of material. • Plastic flow direction different from the forging direction indicated.

  20. Magnetic Barkhausen noise and magneto acoustic emission in pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Neyra Astudillo, Miriam Rocío, E-mail: neyra@cnea.gov.ar [IT Sabato, Universidad Nacional de San Martín, UNSAM, Av. General Paz 1499, Buenos Aires (Argentina); Universidad Tecnológica Nacional UTN, Regional Delta, Buenos Aires (Argentina); López Pumarega, María Isabel, E-mail: lopezpum@cnea.gov.ar [Comisión Nacional de Energía Atómica, CNEA, Av. General Paz 1499, Buenos Aires (Argentina); Núñez, Nicolás Marcelo, E-mail: nnunez@cnea.gov.ar [Comisión Nacional de Energía Atómica, CNEA, Av. General Paz 1499, Buenos Aires (Argentina); Pochettino, Alberto, E-mail: alberto.poch@gmail.com [Comisión Nacional de Energía Atómica, CNEA, Av. General Paz 1499, Buenos Aires (Argentina); Instituto de Investigación e Ingeniería Ambiental (3iA), Campus Miguelete, UNSAM, Av. 25 de Mayo y Francia, 1650 San Martín Argentina (Argentina); Ruzzante, José, E-mail: ruzzante@gmail.com [Universidad Tecnológica Nacional UTN, Regional Delta, Buenos Aires (Argentina); Universidad Nacional de Tres de Febrero UNTREF, Caseros, Buenos Aires (Argentina); Universidad Nacional de Chilecito, UNdeC, La Rioja (Argentina)

    2017-03-15

    Magnetic Barkhausen Noise (MBN) and Magneto Acoustic Emission (MAE) were studied in A508 Class II forged steel used for pressure vessels in nuclear power stations. The magnetic experimental determinations were completed with a macro graphic study of sulfides and the texture analysis of the material. The analysis of these results allows us to determine connections between the magnetic anisotropy, texture and microstructure of the material. Results clearly suggest that the plastic flow direction is different from the forging direction indicated by the material supplier - Highlights: • MBN and MAE studied in nuclear power pressure vessel steel. • Comparison with macro graphic study of sulfides and texture analysis of the material. • Connections with magnetic anisotropy, texture and microstructure of material. • Plastic flow direction different from the forging direction indicated.

  1. Investigation of irradiation embrittlement and annealing behaviour of JRQ pressure vessel steel by instrumented impact tests

    Energy Technology Data Exchange (ETDEWEB)

    Valo, M; Rintamaa, R; Nevalainen, M; Wallin, K; Torronen, K [Technical Research Centre of Finland, Espoo (Finland); Tipping, P [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1994-12-31

    Seven series of A533-B type pressure vessel steel specimens irradiated as well as irradiated - annealed - re-irradiated to different fast neutron fluences (up to 5.10{sup 19}/cm{sup 2}) have been tested with a new type of instrumented impact test machine. The radiation embrittlement and the effect of the intermediate annealing was assessed by using the ductile and cleavage fracture initiation toughness. Although the ductile fracture initiation toughness exhibited scatter, the transition temperature shift corresponding to the dynamic cleavage fracture initiation agreed well with the 41 J Charpy-V shift. The results indicate that annealing is beneficial in restoring mechanical properties in an irradiated nuclear pressure vessel steel. (authors). 8 refs., 11 figs., 1 tab.

  2. Investigation of neutron irradiated reactor vessel steels using post-irradiation annealing techniques

    Energy Technology Data Exchange (ETDEWEB)

    Nakata, Hayato; Fukuya, Koji [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    The matrix damage is known to be a major factor that contributes to embrittlement and hardening of irradiated reactor vessel steels, and is assumed to be composed of the point defect clusters. However field emission gun scanning transmission electron microscopy (FEGSTEM) and atom probe (AP) could not detect any evidence of the matrix damage. In this study, post irradiation annealing experiments combining positron annihilation lineshape analysis (PALA) and hardness experiments were applied to an actual surveillance test specimen and a sample of reactor vessel steel irradiated in a material test reactor (MTR), in order to investigate the matrix damage recovery behavior and its contribution to hardening. It was confirmed that higher fluence increased the hardness and the volume fraction of open volume defects and that higher flux decreased the thermal stability of matrix damage and the effect on hardening. The contribution of matrix damage to hardening could be estimated to be below 30%. (author)

  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.

    1996-01-01

    Embrittlement of reactor pressure vessels (RPVs) can be reduced by thermal annealing at temperatures higher than the normal operating conditions. Although such an annealing process has not been applied to any commercial plants in the United States, one US Army reactor, the BR3 plant in Belgium, and several plants in eastern Europe have been successfully annealed. All available Charpy annealing data were collected and analyzed in this project to develop quantitative models for estimating the recovery in 30 ft-lb (41 J) Charpy transition temperature and Charpy upper shelf energy over a range of potential annealing conditions. Pattern recognition, transformation analysis, residual studies, and the current understanding of the mechanisms involved in the annealing process were used to guide the selection of the most sensitive variables and correlating parameters and to determine the optimal functional forms for fitting the data. The resulting models were fitted by nonlinear least squares. The use of advanced tools, the larger data base now available, and insight from surrogate hardness data produced improved models for quantitative evaluation of the effects of annealing. The quality of models fitted in this project was evaluated by considering both the Charpy annealing data used for fitting and the surrogate hardness data base. The standard errors of the resulting recovery models relative to calibration data are comparable to the uncertainty in unirradiated Charpy data. This work also demonstrates that microhardness recovery is a good surrogate for transition temperature shift recovery and that there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

  4. Reactor pressure vessel steels ASTM A533B and A508 Cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Kemppainen, M.; Toerroenen, K.

    1979-11-01

    This report presents the tensile test results of steels ASTM A533B and A508 Cl.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of the structural integrity of the reactor pressure vessels. The tensile properties were studied between -196 and 300 degC varying austenitizing and tempering temperatures and having two different carbon contents for the heats of A533B. (author)

  5. Reactor pressure vessel steels[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-07-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain 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 a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use.

  6. Analysis of a molten pool natural convection in the APR1400 RPV at a severe accident

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Park, Rae Joon; Kim, Sang Baik

    2005-01-01

    During a hypothetical severe accident, reactor fuel rods and structures supporting them are melted and relocated in the lower head of the reactor vessel. These relocated molten materials could be separated by their density difference and construct metal/oxide stratified pools in the lower head. A decay heat generated from the fuel material is transferred to the vessel wall and upper structures remaining in the reactor vessel by natural convection. As shown in Fig. 1 two-layered stratified molten pool is developed in the reactor lower vessel. The oxidic pool usually constructed by the mixture of uranium oxide and zirconium oxide. The melting temperature of the oxidic material is very high compared to the steel vessel and metallic layer. And highly turbulent natural convection generated by the decay heat enhances heat transfer to the boundary of the oxidic pool. By this thermal mechanism, oxide curst is developed around the oxidic layer as shown in Fig. 1. The oxidic pool is bounded thermally and fluid-dynamically by the developed crust. By this boundedness, the heat transfer structure in the stratified oxidic/metallic pool can be solved separately. The thermal boundary condition of the oxidic pool is isothermal with constant melting temperature of the oxidic material. The decay heat is transfer to side wall and upper interface between oxidic and metallic layer. Turbulent natural convection is dominant heat transfer mechanism in the oxidic pool. The heat transferred from the bottom oxidic layer is imposed to the upper metallic layer. This transferred heat in the metallic pool is removed through side and upper surface, which is augmented also by natural convection developed in the pool. In this study, a molten pool natural convection in the APR1400 RPV during a severe accident is simulated using the Lilac code and the calculated heat flux distribution on the reactor vessel wall is compared with a lumped-parameter (LP) prediction

  7. Effect of heterogeneities on the thermoelectric power of pressure vessel steel

    International Nuclear Information System (INIS)

    Simonet, L.

    2006-12-01

    In service working conditions, the vessel of the Pressurized Water Reactors (PWR) undergoes an ageing due to irradiation. In order to follow the evolution of the mechanical characteristics of the steel in service, EDF launched a surveillance program which consists to carry out mechanical tests on samples aged in reactor. However, the results of these tests have the disadvantage to be affected by the presence of heterogeneities within the steel. Indeed, because of its manufacturing process, the steel contains segregated areas. Thus, EDF launched Thermoelectric Power Measurements (TEP) on the resilience samples of the surveillance program, to complete the mechanical tests and to help with their interpretation. However, these measurements are today difficult to analyse because they include at the same time the effect of the irradiation and the effect of the metallurgical heterogeneities. The aim of this work consisted in evaluating the effect of the heterogeneities on the TEP of the non-irradiated vessel steel. For that, a numerical model was developed which allows to calculate the TEP of a composite structure. We have shown that the model is pertinent to highlight the effect of the heterogeneities on the TEP of the vessel steel, which is considered like a 'matrix'/'segregation' composite. The model allowed us to put emphasis on the influence of different parameters on the TEP measurement. We have thus showed that the measurements conditions have an important effect on the obtained TEP value (influence of the applied pressure, the position of the sample on the device, the site of the metallurgical heterogeneities,...). (author)

  8. Fracture toughness and crack growth resistance of pressure vessel plate and weld metal steels

    International Nuclear Information System (INIS)

    Moskovic, R.

    1988-01-01

    Compact tension specimens were used to measure the initiation fracture toughness and crack growth resistance of pressure vessel steel plates and submerged arc weld metal. Plate test specimens were manufactured from four different casts of steel comprising: aluminium killed C-Mn-Mo-Cu and C-Mn steel and two silicon killed C-Mn steels. Unionmelt No. 2 weld metal test specimens were extracted from welds of double V butt geometry having either the C-Mn-Mo-Cu steel (three weld joints) or one particular silicon killed C-Mn steel (two weld joints) as parent plate. A multiple specimen test technique was used to obtain crack growth data which were analysed by simple linear regression to determine the crack growth resistance lines and to derive the initiation fracture toughness values for each test temperature. These regression lines were highly scattered with respect to temperature and it was very difficult to determine precisely the temperature dependence of the initiation fracture toughness and crack growth resistance. The data were re-analysed, using a multiple linear regression method, to obtain a relationship between the materials' crack growth resistance and toughness, and the principal independent variables (temperature, crack growth, weld joint code and strain ageing). (author)

  9. Stress corrosion cracking studies on ferritic low alloy pressure vessel steel - water chemistry and modelling aspects

    International Nuclear Information System (INIS)

    Tipping, P.; Ineichen, U.; Cripps, R.

    1994-01-01

    The susceptibility of low alloy ferritic pressure vessel steels (A533-B type) to stress corrosion cracking (SCC) degradation has been examined using various BWR type coolant chemistries. Fatigue pre-cracked wedge-loaded double cantilever beams and also constantly loaded 25 mm thick compact tension specimens have shown classical SCC attack. The influence of parameters such as dissolved oxygen content, water impurity level and conductivity, material chemical composition (sulphur content) and stress intensity level are discussed. The relevance of SCC as a life-limiting degradation mechanism for low alloy ferritic nuclear power plant PV steel is examined. Some parameters, thought to be relevant for modelling SCC processes in low alloy steels in simulated BWR-type coolant, are discussed. 8 refs., 1 fig., 4 tabs

  10. Low cycle fatigue behavior of a ferritic reactor pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Apu, E-mail: asarkar@barc.gov.in; Kumawat, Bhupendra K.; Chakravartty, J.K.

    2015-07-15

    The cyclic stress–strain response and the low cycle fatigue (LCF) behavior of 20MnMoNi55 pressure vessel steel were studied. Tensile strength and LCF properties were examined at room temperature (RT) using specimens cut from rolling direction of a rolled block. The fully reversed strain-controlled LCF tests were conducted at a constant total strain rate with different axial strain amplitude levels. The cyclic strain–stress relationships and the strain–life relationships were obtained through the test results, and related LCF parameters of the steel were calculated. The studied steel exhibits cyclic softening behavior. Furthermore, analysis of stabilized hysteresis loops showed that the steel exhibits non-Masing behavior. Complementary scanning electron microscopy examinations were also carried out on fracture surfaces to reveal dominant damage mechanisms during crack initiation, propagation and fracture. Multiple crack initiation sites were observed on the fracture surface. The investigated LCF behavior can provide reference for pressure vessel life assessment and fracture mechanisms analysis.

  11. Isothermal and thermal–mechanical fatigue of VVER-440 reactor pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Fekete, Balazs, E-mail: fekete.mm.bme@gmail.com [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary); Department of Applied Mechanics, Budapest University of Technology and Economics, Muegyetem 5, Budapest H-1111 (Hungary); Trampus, Peter [College of Dunaujvaros, Tancsics 1A, Dunaujvaros H-2400 (Hungary)

    2015-09-15

    Highlights: • We aimed to determine the thermomechanical behaviour of VVER reactor steels. • Material tests were developed and performed on GLEEBLE 3800 physical simulator. • Coffin–Manson curves and parameters were derived. • High accuracy of the strain energy based evaluation was found. • The observed dislocation evolution correlates with the mechanical behaviour. - Abstract: The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin–Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  12. Neutron irradiation embrittlement of reactor pressure vessel steel 20 MnMoNi55 weld

    International Nuclear Information System (INIS)

    Ghoneim, M.M.

    1987-05-01

    The effect of neutron irradiation on the mechanical and fracture properties of an 'improved' 20 MnMoNi 55 Pressure Vessel Steel (PVS) weld was investigated. In addition to very low residual element content, especially Cu (0.035 wt.%), and relatively higher Ni content (0.9 wt.%), this steel has higher strength (30% more) than the steels used currently in nuclear reactor pressure vessels. The material was irradiated to 3.5x10 19 and 7x10 19 n/cm 2 (E > 1 Mev) at 290 0 C and 2.5x10 19 n/cm 2 (E > 1 MeV) at 160 0 C in FRJ-1 and FRJ-2 research reactors at KFA, Juelich, F.R.G. Test methods used in the evaluation included instrumented impact testing of standard and precracked Charpy specimens, tensile, and fracture toughness testing. Instrumented impact testing provided load and energy vs. time (deflection) data in addition to energy absorption data. The results indicated that the investigated high strength improved steel is more resistant to irradiation induced embrittlement than conventional PVSs. (orig./IHOE)

  13. The irradiation embrittlement of two pressure vessel steels -Contribution of local approach

    Energy Technology Data Exchange (ETDEWEB)

    Soulat, P; Marini, B [CEA Centre d` Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France). Service de Recherches Metallurgiques Appliquees; Miannay, D; Horowitz, H [CEA Centre d` Etudes de Fontenay-aux-Roses, 92 (France). Inst. de Protection et de Surete Nucleaire; Schill, R [CEA Centre d` Etudes de Saclay, 91 - Gif-sur-Yvette (France). Dept. de Mecanique et de Technologie

    1994-12-31

    Within the IAEA Coordinated Research Programme on ``Optimizing the Reactor Pressure Vessel Surveillance Programmes and their Analyses``, the French participation has been focused on the contribution of the local approach to the determination of the sensitivity to radiation embrittlement of two different pressure vessel steels: a low sensitive French forging steel (FFA) and a high sensitive ``monitor`` Japanese plate steel (JRQ) were irradiated to a fluence of 3.10{sup 19} n/cm{sup 2} at 290 C. The irradiation embrittlement of the two steels measured by the shift of Charpy V transition curves is in good agreement with the estimated shifts given by theoretical prediction. The fracture toughness properties were examined at low temperature with brittle fracture, and at service temperature (290 C), with ductile tearing. The values of K{sub 1C} or K{sub JC} for the brittle fracture and J{sub 1C} for the ductile fracture are compared to predictions established using the local approach of cleavage fracture (Weibull analysis) and the critical rate of void growth respectively. 8 refs., 14 figs., 10 tabs.

  14. Acoustic emission during the elastic-plastic deformation of low alloy reactor pressure vessel steels. I

    International Nuclear Information System (INIS)

    Holt, J.; Goddard, D.J.

    1980-01-01

    Measurements of the acoustic emission behaviour of A533B and C-Mn low alloy reactor pressure vessel steels subjected to uniaxial tensile deformation are described. The effects on the emission activity of the rolling plane orientation and the carbide morphology were examined. Detailed discussions are given of the stress dependence of the emission activity below yield and of its recovery by annealing at the stress relief temperature. It is shown that the dominant emission source is the same in both steels and is associated with inclusions, such as MnS, elongated by the rolling process, the carbide morphology being relatively unimportant. A criterion for the occurrence of an emission is obtained which is directly analogous to the general criterion for yielding. It is also shown that a large fraction, at least, of the emission activity arises from a recoverable process such as localized yielding around inclusions or limited inclusion decohesion and not from inclusion fracture. Low activity in C-Mn steel taken from reactor pressure vessels, previously attributed to spheroidization of carbides, is shown to be due to the limited acoustic recovery of these relatively high sulphur content steels when annealed at the stress relief temperature. It is concluded that the limited amplitudes of these emissions during deformation severely restrict their potential application in practice. (Auth.)

  15. 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

  16. 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)

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, Masayuki; Nagai, Yasuyoshi; Tang, Zheng; Yubuta, Kunio [Tohoku Univ., Sendai (Japan). Inst. for Materials Research; Suzuki, Masahide [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    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) in a Fe-Cu single crystal and was agreed well with that from a band structure calculation. Theoretical calculation of positron confinement in Fe-Cu model alloys showed that a positron quantum dot state induced by positron affinity is attained for the embedded precipitates larger than 1 nm. A new position sensitive detector with a function of one dimensional angular correlation of annihilation radiation (1D-ACAR) has been developed that enables high resolution experiments over wide ranges of momentum distribution. (author)

  17. Models for 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.

    1995-05-01

    The reactor pressure vessel (RPV) surrounding the core of a commercial nuclear power plant is subject to embrittlement due to exposure to high energy neutrons. The effects of irradiation embrittlement can be reduced by thermal annealing at temperatures higher than the normal operating conditions. However, a means of quantitatively assessing the effectiveness of annealing for embrittlement recovery is needed. The objective of this work was to analyze the pertinent data on this issue 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. Data were gathered from the Test Reactor Embrittlement Data Base and from various annealing reports. An analysis data base was developed, reviewed for completeness and accuracy, and documented as part of this work. Independent variables considered in the analysis included material chemistries, annealing time and temperature, irradiation time and temperature, fluence, and flux. To identify important variables and functional forms for predicting embrittlement recovery, advanced statistical techniques, including pattern recognition and transformation analysis, were applied together with current understanding of the mechanisms governing embrittlement and recovery. Models were calibrated using multivariable surface-fitting techniques. Several iterations of model calibration, evaluation with respect to mechanistic and statistical considerations, and comparison with the trends in hardness data produced correlation models for estimating Charpy upper shelf energy and transition temperature after irradiation and annealing. This work provides a clear demonstration that (1) microhardness recovery is generally a very good surrogate for shift recovery, and (2) there is a high level of consistency between the observed annealing trends and fundamental models of embrittlement and recovery processes

  18. Application of small specimens to fracture mechanics characterization of irradiated pressure vessel steels

    International Nuclear Information System (INIS)

    Sokolov, M.A.; Wallin, K.; McCabe, D.E.

    1996-01-01

    In this study, precracked Charpy V-notch (PCVN) specimens were used to characterize the fracture toughness of unirradiated and irradiated reactor pressure vessel steels in the transition region by means of three-point static bending. Fracture toughness at cleavage instability was calculated in terms of elastic-plastic K Jc values. A statistical size correction based upon weakest-link theory was performed. The concept of a master curve was applied to analyze fracture toughness properties. Initially, size-corrected PCVN data from A 533 grade B steel, designated HSST Plate O2, were used to position the master curve and a 5% tolerance bound for K Jc data. By converting PCVN data to IT compact specimen equivalent K Jc data, the same master curve and 5% tolerance bound curve were plotted against the Electric Power Research Institute valid linear-elastic K Jc database and the ASME lower bound K Ic curve. Comparison shows that the master curve positioned by testing several PCVN specimens describes very well the massive fracture toughness database of large specimens. These results give strong support to the validity of K Jc with respect to K Ic in general and to the applicability of PCVN specimens to measure fracture toughness of reactor vessel steels in particular. Finally, irradiated PCVN specimens of other materials were tested, and the results are compared to compact specimen data. The current results show that PCVNs demonstrate very good capacity for fracture toughness characterization of reactor pressure vessel steels. It provides an opportunity for direct measurement of fracture toughness of irradiated materials by means of precracking and testing Charpy specimens from surveillance capsules. However, size limits based on constraint theory restrict the operational test temperature range for K Jc data from PCVN specimens. 13 refs., 8 figs., 1 tab

  19. Probabilistic Assessment of the Design and Safety of HSLA-100 Steel Confinement Vessels

    Energy Technology Data Exchange (ETDEWEB)

    R.M. Dolin

    2003-03-03

    This probabilistic approach for assessing the design and safety of the HSLA-100 steel confinement vessel used for a DynEx test involved the probability of failure for several scenarios, in which a fragment may penetrate the vessel. The samples involve vessel thicknesses of 1 inch, 2 inches, and 5.25 inches--the combined thicknesses of the 2 inch containment vessel and the 3.25 inch safety vessel. Two simulation approaches were used for each scenario to assess the probability of failure. The Likelihood of Occurrence method simultaneously models all likely fragment events of a test, for which the net probability of failure is the sum of all the fragment events. The Stochastic Sampling method determines the probability of a fragment perforation on the basis of a logical model and takes the overall probability that an experiment results in failure as the maximum probability for any fragment event. With margin and safety assessments taken into account, it was concluded that the one and two inch thicknesses by themselves are inadequate for containing a DynEx test. The 5.25 inch thickness was determined to be safe by the Likelihood of Occurrence method and nearly adequate by the Stochastic Sampling simulation.

  20. Isothermal and thermal-mechanical fatigue of VVER-440 reactor pressure vessel steels

    Science.gov (United States)

    Fekete, Balazs; Trampus, Peter

    2015-09-01

    The fatigue life of the structural materials 15Ch2MFA (CrMoV-alloyed ferritic steel) and 08Ch18N10T (CrNi-alloyed austenitic steel) of VVER-440 reactor pressure vessel under completely reserved total strain controlled low cycle fatigue tests were investigated. An advanced test facility was developed for GLEEBLE-3800 physical simulator which was able to perform thermomechanical fatigue experiments under in-service conditions of VVER nuclear reactors. The low cycle fatigue results were evaluated with the plastic strain based Coffin-Manson law, and plastic strain energy based model as well. It was shown that both methods are able to predict the fatigue life of reactor pressure vessel steels accurately. Interrupted fatigue tests were also carried out to investigate the kinetic of the fatigue evolution of the materials. On these samples microstructural evaluation by TEM was performed. The investigated low cycle fatigue behavior can provide reference for remaining life assessment and lifetime extension analysis.

  1. Formation of radiation induced precipitates in VVER RPV materials

    International Nuclear Information System (INIS)

    Platonov, P.A.; Chernobaeva, A.A.

    2016-01-01

    This paper presents an analysis of experimental results received in course of research of copper-enriched precipitates (Cu-precipitates) and nickel-manganese-silicon clusters (Ni-Mn-Si clusters), which are formed in steels of VVER-type reactor pressure vessels (RPVs) under neutron irradiation. Based on this analysis, a hypothetical model is suggested for cluster formation in course of evolution of a cascade region. The model presumes cluster formation in two stages. At the first stage, in course of cascade region crystallization, a stable cluster is formed in the center of the cascade region, which consists of vacancies and Cu atoms following the mechanism of the inverse Kirkendall effect. At the second stage, diffusion of Ni, Mn and P atoms with a flow of vacancies from the matrix takes place to form a cluster. The size of a cluster is limited by a balance of vacancies' flows entering and leaving the cluster. The paper also considers a possibility of stabilization of atomic-vacancy cluster due to uneven distribution of Ni, Mn and P atoms, which explains dependence of cluster density on the content of these elements. Kinetics of cluster formation and evolution presumed by suggested model is analyzed. It is demonstrated that a fall in cluster density and an increase in their size under high irradiation doses may be caused by a decrease of matrix supersaturation with vacancies resulting from high density of dislocation loops. - Highlights: • The analysis of the mechanism of formation of radiation-induced clusters in RPV steels has been done. • Radiation-induced clusters are formed after the mechanism based on the inverse Kirkendall effect in two stages. • At post-dynamic stage a flow of vacancies moving to the center of the cascade entrains Cu atoms contained and forms a stable atom-vacancies cluster. • At the 2nd stage Cu, Ni, Mn, Si atoms forming complexes with vacancies diffuse into a cluster driving out Fe and Cr atoms from the cluster. • The cluster

  2. Heavy-section steel irradiation program. Semiannual progress report, September 1993--March 1994

    International Nuclear Information System (INIS)

    Corwin, W.R.

    1995-04-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only component in the primary pressure boundary for which, if it should rupture, the engineering safety systems cannot assure protection from core damage. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. In particular, ft is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. The Heavy-Section Steel (HSS) Irradiation Program has been established; its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties of typical pressure-vessel steels, as they relate to light-water RPV integrity. The program includes the direct continuation of irradiation studies previously conducted within the HSS Technology Program augmented by enhanced examinations of the accompanying microstructural changes. During this period, the report on the duplex-type crack-arrest specimen tests from Phase 11 of the K la program was issued, and final preparations for testing the large, irradiated crack-arrest specimens from the Italian Committee for Research and Development of Nuclear Energy and Alternative Energies were completed. Tests on undersize Charpy V-notch (CVN) energy specimens in the irradiated and annealed weld 73W were completed. The results are described in detail in a draft NUREG report. In addition, the ORNL investigation of the embrittlement of the High Flux Isotope RPV indicated that an unusually large ratio of the high-energy gamma-ray flux to fast-neutron flux is most likely responsible for the apparently accelerated embrittlement

  3. Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

    1993-01-01

    In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

  4. Microstructural characterization of atom clusters in irradiated pressure vessel steels and model alloys

    Energy Technology Data Exchange (ETDEWEB)

    Auger, P; Pareige, P [Rouen Univ., 76 - Mont-Saint-Aignan (France); Akamatsu, M; Van Duysen, J C [Electricite de France (EDF), 77 - Ecuelles (France)

    1994-12-31

    In order to characterize the microstructural evolution of iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions, and, for comparison, low copper model alloys irradiated with neutrons and electrons, have been studied through small angle neutron scattering and atom probe experiments. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex; solute atoms such as Ni, Mn and Si, sometimes associated with Cu, segregate as ``clouds`` more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs.

  5. Characterisation of creep cavitation damage in a stainless steel pressure vessel using small angle neutron scattering

    CERN Document Server

    Bouchard, P J; Treimer, W

    2002-01-01

    Grain-boundary cavitation is the dominant failure mode associated with initiation of reheat cracking, which has been widely observed in austenitic stainless steel pressure vessels operating at temperatures within the creep range (>450 C). Small angle neutron scattering (SANS) experiments at the LLB PAXE instrument (Saclay) and the V12 double-crystal diffractometer of the HMI-BENSC facility (Berlin) are used to characterise cavitation damage (in the size range R=10-2000 nm) in a variety of creep specimens extracted from ex-service plant. Factors that affect the evolution of cavities and the cavity-size distribution are discussed. The results demonstrate that SANS techniques have the potential to quantify the development of creep damage in type-316H stainless steel, and thereby link microstructural damage with ductility-exhaustion models of reheat cracking. (orig.)

  6. Characterization of atom clusters in irradiated pressure vessel steels and model alloys

    International Nuclear Information System (INIS)

    Auger, P.; Pareige, P.; Akamatsu, M.; Van Duysen, J.C.

    1993-12-01

    In order to characterize the microstructural evolution of the iron solid solution under irradiation, two pressure vessel steels irradiated in service conditions and, for comparison, low copper model alloys irradiated with neutrons and electrons have been studied. The characterization has been carried out mainly thanks to small angle neutron scattering and atom probe experiments. Both techniques lead to the conclusion that clusters develop with irradiations. In Fe-Cu model alloys, copper clusters are formed containing uncertain proportions of iron. In the low copper industrial steels, the feature is more complex. Solute atoms like Ni, Mn and Si, sometimes associated with Cu, segregate as ''clouds'' more or less condensed in the iron solid solution. These silicides, or at least Si, Ni, Mn association, may facilitate the copper segregation although the initial iron matrix contains a low copper concentration. (authors). 24 refs., 3 figs., 2 tabs

  7. Re-austenitisation of chromium-bearing pressure vessel steels during the weld thermal cycle

    International Nuclear Information System (INIS)

    Dunne, Druce; Li, Huijun; Jones, Christopher

    2013-01-01

    Steels with chromium contents between 0.5 and 12 wt% are commonly used for fabrication of creep resistant pressure vessels (PV) for the power generation industry. Most of these steels are susceptible to Type IV creep failure in the intercritical and/ or grain refined regions of the heat affected zone (HAZ) of the parent metal. The re-austenitisation process plays a central role in establishing the transformed microstructures and the creep resistance of the various sub-zones of the HAZ. The high alloy content and the presence of alloy-rich carbides in the as-supplied parent plate can significantly retard the kinetics of transformation to austenite, resulting in both incomplete austenitisation and inhomogeneous austenite. Overlapping weld thermal cycles in multi-pass welds add further complexity to the progressive development of microstructure over the course of the welding process. In order to clarify structural evolution, thermal simulation has been used to study the effects of successive thermal cycles on the structures and properties of the HAZ of 2.25Cr-1Mo steel. The results showed that, before post-weld heat treatment (PWHT), the HAZ microstructures and properties, particularly in doubly reheated sub-zones, were highly heterogeneous and differed markedly from those of the base steel. It is concluded that close control of the thermal cycle by pre-heat, weld heat input and post-heat is necessary to obtain a heat affected zone with microstructures and properties compatible with those of the base plate.

  8. Evolution of manganese–nickel–silicon-dominated phases in highly irradiated reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Wells, Peter B.; Yamamoto, Takuya; Miller, Brandon; Milot, Tim; Cole, James; Wu, Yuan; Odette, G. Robert

    2014-01-01

    Formation of a high density of Mn–Ni–Si nanoscale precipitates in irradiated Cu-free and Cu-bearing reactor pressure vessel steels could lead to severe unexpected embrittlement. Models long ago predicted that these precipitates, which are not treated in current embrittlement prediction models, would emerge only at high fluence. However, the mechanisms and variables that control Mn–Ni–Si precipitate formation, and their detailed characteristics, have not been well understood. High flux irradiations of six steels with systematic variations in Cu and Ni contents were carried out at ∼295 °C to high and very high neutron fluences of ∼1.3 × 10 20 and ∼1.1 × 10 21 n cm −2 . Atom probe tomography shows that significant mole fractions of Mn–Ni–Si-dominated precipitates form in the Cu-bearing steels at ∼1.3 × 10 20 n cm −2 , while they are only beginning to develop in Cu-free steels. However, large mole fractions of these precipitates, far in excess of those found in previous studies, are observed at 1.1 × 10 21 n cm −2 at all Cu contents. At the highest fluence, the precipitate mole fractions primarily depend on the alloy Ni, rather than Cu, content. The Mn–Ni–Si precipitates lead to very large increases in measured hardness, corresponding to yield strength elevations of up to almost 700 MPa

  9. Heavy-section steel irradiation program. Semiannual progress report, October 1996--March 1997

    International Nuclear Information System (INIS)

    Rosseel, T.M.

    1998-02-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. Because the RPV is the only key safety-related component of the plant for which a redundant backup system does not exist, it is imperative to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance that occurs during service. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established. Its primary goal is to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels as they relate to light-water RPV integrity. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into eight tasks: (1) program management, (2) irradiation effects in engineering materials, (3) annealing, (4) microstructural analysis of radiation effects, (5) in-service irradiated and aged material evaluations, (6) fracture toughness curve shift method, (7) special technical assistance, and (8) foreign research interactions. The work is performed by the Oak Ridge National Laboratory

  10. Lower Length Scale Model Development for Embrittlement of Reactor Presure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schwen, Daniel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Chakraborty, Pritam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, Xianming [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the lower-length-scale effort during FY 2016 in developing mesoscale capabilities for microstructure evolution, plasticity and fracture in reactor pressure vessel steels. During operation, reactor pressure vessels are subject to hardening and embrittlement caused by irradiation induced defect accumulation and irradiation enhanced solute precipitation. Both defect production and solute precipitation start from the atomic scale, and manifest their eventual effects as degradation in engineering scale properties. To predict the property degradation, multiscale modeling and simulation are needed to deal with the microstructure evolution, and to link the microstructure feature to material properties. In this report, the development of mesoscale capabilities for defect accumulation and solute precipitation are summarized. A crystal plasticity model to capture defect-dislocation interaction and a damage model for cleavage micro-crack propagation is also provided.

  11. Effects of temperature on corrosion fatigue crack growth of pressure vessel steels in PWR coolant

    International Nuclear Information System (INIS)

    Tice, D.R.; Bramwell, I.L.; Fairbrother, H.; Worswick, D.

    1994-01-01

    This paper presents experimental results concerning crack propagation rates in A508-III pressure vessel steel (medium sulphur content) exposed to PWR primary water at temperatures between 130 and 290 C. The results indicate that the greatest increase in corrosion fatigue crack growth rate occurs at temperatures in the range 150 to 200 C. Under these conditions, there was a marked change in the appearance of the fracture surface, with extensive micro-branching of the crack front and occasional bifurcation of the whole crack path. In contrast, at 290 C, the fracture surface is smoother, similar to that due to inert fatigue. The implication of these observations for assessment of the pressure vessel integrity, is examined. 14 refs., 15 figs., 3 tabs

  12. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Rui Wu; Seitisleam, F.; Sandstroem, R. [Swedish Institute for Metals Research, Stockholm (Sweden)

    1998-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  13. Guide to the periodic inspection of nuclear reactor steel pressure vessels

    International Nuclear Information System (INIS)

    1969-01-01

    This Guide is intended to provide general information and guidance to reactor owners or operators, inspection authorities, certifying authorities or regulatory bodies who are responsible for establishing inspection procedures for specific reactors or reactor types, and for the preparation of national codes or standards. The recommendations of the Guide apply primarily to water-cooled steel reactor vessels which are at a sufficiently early stage of design so that recommendations to provide accessibility for inspection can be incorporated into the early stages of design and inspection planning. However, much of the contents of the Guide are also applicable in part to vessels for other reactor types, such as gas-cooled, pressure-tube, or liquid-metal-cooled reactors, and also to some existing water-cooled reactors and reactors which are in advanced stage of design or construction. 46 refs, figs, 1 tab

  14. Creep crack growth in a reactor pressure vessel steel at 360 deg C

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Rui; Seitisleam, F; Sandstroem, R [Swedish Institute for Metals Research, Stockholm (Sweden)

    1999-12-31

    Plain creep (PC) and creep crack growth (CCG) tests at 360 deg C and post metallography were carried out on a low alloy reactor pressure vessel steel (ASTM A508 class 2) with different microstructures. Lives for the CCG tests were shorter than those for the PC tests and this is more pronounced for simulated heat affected zone microstructure than for the parent metal at longer lives. For the CCG tests, after initiation, the cracks grew constantly and intergranularly before they accelerated to approach rupture. The creep crack growth rate is well described by C*. The relations between reference stress, failure time and steady crack growth rate are presented for the CCG tests. It is demonstrated that the failure stress due to CCG is considerably lower than the yield stress at 360 deg C. Consequently, the CCG will control the static strength of a reactor vessel. (orig.) 17 refs.

  15. Underwater cutting of stainless steel plate and pipe for dismantling reactor pressure vessels

    International Nuclear Information System (INIS)

    Hamasaki, M.; Tateiwa, F.; Kanatani, F.; Yamashita, S.

    1982-01-01

    A consumable electrode water jet cutting technique is described. Satisfactory underwater cutting of 80mm stainless steel plate using a current of 2000A and at a water depth of 200mm has been demonstrated. The electrical requirements for this arc welding method applied to cutting were found to be approximately one third those required for conventional plasma arc cutting for the same thickness plate. An application of this technique might be found in the dismantling of atomic reactor pressure vessels, and parts of commercial atomic reactors. (author)

  16. Temperature dependence of the fracture toughness and the cleavage fracture strength of a pressure vessel steel

    International Nuclear Information System (INIS)

    Kotilainen, H.

    1980-01-01

    A new model for the temperature dependence of the fracture toughness has been sought. It is based on the yielding processes at the crack tip, which are thought to be competitive with fracture. Using this method a good correlation between measured and calculated values of fracture toughness has been found for a Cr-Mo-V pressure vessel steel as well as for A533B. It has been thought that the application of this method can reduce the number of surveillance specimens in nuclear reactors. A method for the determination of the cleavage fracture strength has been proposed. 28 refs

  17. Strain ageing of nuclear pressure vessel steels A533B and A508 cl.2

    International Nuclear Information System (INIS)

    Pelli, R.; Toerroenen, K.

    1978-04-01

    The susceptibility of the reactor pressure vessel steels A533B and A508 cl.2 to strain ageing has been studied using conventional tensile and impact testing of prestrained and aged specimens. The results show a modest susceptibility, seen as an increase in yield strength and Charpy V transition temperatures. The effect of varying alloying additions within the range of normal production was not observed, but the initial mechanical properties clearly affect the strain ageing. The lower the initial yield strength, the higher increase in strength and the lower increase in transition temperature is observed. (author)

  18. Revision of the fracture models in steels for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Darwish, F A.I. [Pontificia Univ. Catolica do Rio de Janeiro (Brazil). Dept. de Ciencia dos Materiais e Metalurgia

    1981-01-01

    The variation of toughness with the temperature of steels used in the fabrication of nuclear pressure vessels is presented and discuted by mathematical models aiming to reach a critical value of stress or deformation at the moment of the fracture. The mathematical model considered are compatible with the fracture micromechanisms in action and they are capable of foreseeing the variations in the toughness from the mechanical properties evaluated in the tension test. The neutron irradiation effects in the toughness as well as in the variation of this toughness with the operating temperature are still described.

  19. 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

  20. Investigation on multilayer failure mechanism of RPV with a high temperature gradient from core meltdown scenario

    Energy Technology Data Exchange (ETDEWEB)

    Jianfeng, Mao, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China); Xiangqing, Li [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Shiyi, Bao, E-mail: bsy@zjut.edu.cn [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China); Lijia, Luo [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Zengliang, Gao [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China)

    2016-12-15

    Highlights: • The multilayer failure mechanism is investigated for RPV under CHF. • Failure time and location of RPV are predicted under various SA scenarios. • The structural behaviors are analyzed in depth for creep and plasticity. • The effect of internal pressure and temperature gradient is considered. • The structural integrity of RPV is secured within the required 72 creep hours. - Abstract: The Fukushima accident shows that in-vessel retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still exists inside the reactor pressure vessel (RPV). In the traditional concept of IVR, the pressure is supposed to successfully be released, and the temperature distributed among the wall thickness is assumed to be uniform. However, this concept is seriously challenged by reality of Fukushima accident with regard to the existence of both internal pressure and high temperature gradient. Therefore, in order to make the IVR mitigation strategy succeed, the numerical investigation of the lower head behavior and its failure has been performed for several internal pressures under high temperature gradient. According to some requirements in severe accident (SA) management of RPV, it should be ensured that the IVR mitigation takes effect in preventing the failure of the structure within a period of 72 h. Subsequently, the failure time and location have to be predicted under the critical heat flux (CHF) loading condition for lower head, since the CHF is limit thermal boundary before the melt-through of RPV. In illustrating the so called ‘multilayer failure mechanism’, the structural behaviors of RPV are analyzed in terms of the stress, creep strain, deformation, damage on selected paths.

  1. Investigation on multilayer failure mechanism of RPV with a high temperature gradient from core meltdown scenario

    International Nuclear Information System (INIS)

    Jianfeng, Mao; Xiangqing, Li; Shiyi, Bao; Lijia, Luo; Zengliang, Gao

    2016-01-01

    Highlights: • The multilayer failure mechanism is investigated for RPV under CHF. • Failure time and location of RPV are predicted under various SA scenarios. • The structural behaviors are analyzed in depth for creep and plasticity. • The effect of internal pressure and temperature gradient is considered. • The structural integrity of RPV is secured within the required 72 creep hours. - Abstract: The Fukushima accident shows that in-vessel retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still exists inside the reactor pressure vessel (RPV). In the traditional concept of IVR, the pressure is supposed to successfully be released, and the temperature distributed among the wall thickness is assumed to be uniform. However, this concept is seriously challenged by reality of Fukushima accident with regard to the existence of both internal pressure and high temperature gradient. Therefore, in order to make the IVR mitigation strategy succeed, the numerical investigation of the lower head behavior and its failure has been performed for several internal pressures under high temperature gradient. According to some requirements in severe accident (SA) management of RPV, it should be ensured that the IVR mitigation takes effect in preventing the failure of the structure within a period of 72 h. Subsequently, the failure time and location have to be predicted under the critical heat flux (CHF) loading condition for lower head, since the CHF is limit thermal boundary before the melt-through of RPV. In illustrating the so called ‘multilayer failure mechanism’, the structural behaviors of RPV are analyzed in terms of the stress, creep strain, deformation, damage on selected paths.

  2. Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

    1991-01-01

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

  3. Estimating residual life of alloy 600 RPV penetrations

    International Nuclear Information System (INIS)

    Hunt, E.S.; White, G.A.; Pathania, R.; Arey, M.L.; Whitaker, D.E.

    1996-01-01

    Primary water stress corrosion cracking (PWSCC) of Alloy 600 penetrations PWR in reactor pressure vessel (RPV) heads has become a significant economic concern worldwide. PWSCC of these penetrations has led to extended maintenance outages, expensive inspections and repairs, and in some cases, replacement of the entire vessel head. This paper describes methodology developed to predict the remaining life of Alloy 600 penetrations in reactor vessel heads. Predictions of remaining life are an important input to planning models used by utilities to select a strategy for responding to the PWSCC issue at the lowest life cycle cost with an acceptably low risk of leakage. The remaining life of RPV penetrations is determined using the results of inspections of penetrations and statistical methods to predict future degradation. The analysis takes into account the effects of material properties, welding residual stresses, and operating temperature on PWSCC initiation and growth. The probability of developing cracks of various depths is assessed using Monte Carlo methods which provide for uncertainties in the input assumptions. For plants which have not yet performed inspections, remaining life predictions are based on inspection results from similar plants which have performed inspections with corrections made for known differences in design details, material properties and operating conditions

  4. Models for ductile crack initiation and tearing resistance under mode 1 loading in pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, M.R.

    1988-06-01

    Micromechanistic models are presented which aim to predict plane strain ductile initiation toughness, tearing resistance and notched bar fracture strains in pressure vessel steels under monotonically increasing tensile (mode 1) loading. The models for initiation toughness and tearing resistance recognize that ductile fracture proceeds by the growth and linkage of voids with the crack-tip. The models are shown to predict the trend of initiation toughness with inclusion spacing/size ratio and can bound the available experimental data. The model for crack growth can reproduce the tearing resistance of a pressure vessel steel up to and just beyond crack growth initiation. The fracture strains of notched bars pulled in tension are shown to correspond to the achievement of a critical volume fraction of voids. This criterion is combined with the true stress - true strain history of a material point ahead of a blunting crack-tip to predict the initiation toughness. An attempt was made to predict the fracture strains of notched tensile bars by adopting a model which predicts the onset of a shear localization phenomenon. Fracture strains of the correct order are computed only if a ''secondary'' void nucleation event at carbide precipitates is taken into account. (author)

  5. On flux effects in a low alloy steel from a Swedish reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Boåsen, Magnus, E-mail: boasen@kth.se [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Ehrnstén, Ulla [VTT Technical Research Centre of Finland Ltd, PO Box 1000, FI-02044 VTT (Finland)

    2017-02-15

    This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations. - Highlights: • Hardness testing is combined with post irradiation annealing at 330, 360 and 390 °C. • Unstable matrix defects is studied in a reactor pressure vessel steel. • Comparison between surveillance material and accelerated irradiation. • No evidence of unstable matrix defects, i.e. not present in studied material. • Difference in hardness recovery between irradiation conditions found at 390 °C.

  6. Failure prediction of low-carbon steel pressure vessel and cylindrical models

    International Nuclear Information System (INIS)

    Zhang, K.D.; Wang, W.

    1987-01-01

    The failure loads predicted by failure assessment methods (namely the net-section stress criterion; the EPRI engineering approach for elastic-plastic analysis; the CEGB failure assessment route; the modified R6 curve by Milne for strain hardening; and the failure assessment curve based on J estimation by Ainsworth) have been compared with burst test results on externally, axially sharp notched pressure vessel and open-ended cylinder models made from typical low-carbon steel St45 seamless tube which has a transverse true stress-strain curve of straight-line and parabola type and a high value of ultimate strength to yield. It was concluded from the comparison that whilst the net-section stress criterion and the CEGB route did not give conservative predictions, Milne's modified curve did give a conservative and good prediction; Ainsworth's curve gave a fairly conservative prediction; and EPRI solutions also could conditionally give a good prediction but the conditions are still somewhat uncertain. It is suggested that Milne's modified R6 curve is used in failure assessment of low-carbon steel pressure vessels. (author)

  7. Gigacycle fatigue behaviour of austenitic stainless steels used for mercury target vessels

    International Nuclear Information System (INIS)

    Naoe, Takashi; Xiong, Zhihong; Futakawa, Masatoshi

    2016-01-01

    A mercury enclosure vessel for the pulsed spallation neutron source manufactured from a type 316L austenitic stainless steel, a so-called target vessel, suffers the cyclic loading caused by the proton beam induced pressure waves. A design criteria of the JSNS target vessel which is defined based on the irradiation damage is 2500 h at 1 MW with a repetition rate of 25 Hz, that is, the target vessel suffers approximately 10 9 cyclic loading while in operation. Furthermore, strain rate of the beam window of the target vessel reaches 50 s −1 at the maximum, which is much higher than that of the conventional fatigue. Gigacycle fatigue strength up to 10 9 cycles for solution annealed 316L (SA) and cold-worked 316L (CW) were investigated through the ultrasonic fatigue tests. Fatigue tests were performed under room temperature and 250 °C which is the maximum temperature evaluated at the beam window in order to investigate the effect of temperature on fatigue strength of SA and CW 316L. The results showed that the fatigue strength at 250 °C is clearly reduced in comparison with room temperature, regardless of cold work level. In addition, residual strength and microhardness of the fatigue tested specimen were measured to investigate the change in mechanical properties by cyclic loading. Cyclic hardening was observed in both the SA and CW 316L, and cyclic softening was observed in the initial stage of cyclic loading in CW 316L. Furthermore, abrupt temperature rising just before fatigue failure was observed regardless of testing conditions.

  8. A Study on the Coupled FEM-Analysis for Reactor Vessel Lower Head of APR1400 under the Severe Accident Scenario

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyonam; Namgung, Ihn [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    For the stabilization of the RPV the in-vessel retention strategy with external reactor vessel cooling (IVR-ERVC) is adopted in APR1400. Under this severe accident condition, a good understanding of the mechanical behavior of the reactor vessel lower head (RVLH) is necessary both for verification of structural integrity and for improving the design applying appropriate accident mitigation strategies. The purpose of this study is to develop the analysis method of the RVLH with thermo-mechanical analysis using FEM tool (ANSYS v.15) in case of core-melting severe accident condition, and then analyze the RVLH of APR1400 including creep behavior. The plastic strain can be the major cause of lower head failure on the reactor vessel, and the creep cannot be not negligible factor of the failure under the severe accident condition. In the study, we applied constant convection coefficient at assumed temperature on the outside wall of RPV and substitute creep data of SA-508. In addition, it was found that the steel ablation at the interface between corium and vessel steel is not only a thermal phenomenon in the METCOR experiments. Corrosion processes and the formation of eutectics lead to the erosion of the vessel steel at temperatures that are significantly lower than the melting temperature of steel. It called thermo-chemical attack of the corium (corrosion). Reduced wall thickness because of the thermo-chemical effect by corium increase the equivalent plastic strain, and decrease the minimum time to reach 20% creep strain.

  9. Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Kristina, E-mail: kristina.lindgren@chalmers.se [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Boåsen, Magnus [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Stiller, Krystyna [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Vattenfall Ringhals AB, SE-430 22 Väröbacka (Sweden); Thuvander, Mattias [Department of Physics, Chalmers University of Technology, SE-412 96 Göteborg (Sweden)

    2017-05-15

    Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58%) low Cu (0.04%) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance material has a lower cluster number density, but larger clusters. The resulting impact on the mechanical properties of these two effects cancel out, resulting in a measured hardness that seems to be on the same trend as the high flux material. The dispersed barrier hardening model with an obstacle strength factor of 0.15 was found to reproduce the increase in hardness. In the investigated high flux materials, the clusters' Cu content was higher. - Highlights: •Clustering in a low Cu, high Ni reactor pressure vessel steel weld is studied. •The clusters nucleate and grow during irradiation, and consist of Ni, Mn, Si, and Cu. •High flux neutron irradiated material is compared to surveillance material. •High flux was found to result in smaller clusters with a larger number density. •Hardness follows the same dependence on fluence, independent of flux.

  10. Characterization of matrix damage in ion-irradiated reactor vessel steel

    International Nuclear Information System (INIS)

    Fujii, Katsuhiko; Fukuya, Koji

    2004-01-01

    Exact nature of the matrix damage, that is one of radiation-induced nano-scale microstructural features causing radiation embrittlement of reactor vessel, in irradiated commercial steels has not been clarified yet by direct characterization using transmission electron microscopy (TEM). We designed a new preparation method of TEM observation samples and applied it to the direct TEM observation of the matrix damage in the commercial steel samples irradiated by ions. The simulation irradiation was carried out by 3 MeV Ni 2+ ion to a dose of 1 dpa at 290degC. Thin foil specimens for TEM observation were prepared using the modified focused ion beam method. A weak-beam TEM study was carried out for the observation of matrix damage in the samples. Results of this first detailed observation of the matrix damage in the irradiated commercial steel show that it is consisted of small dislocation loops. The observed and analyzed dislocation loops have Burgers vectors b = a , and a mean image size and the number density are 2.5 nm and about 1 x 10 22 m -3 , respectively. In this experiment, all of the observed dislocation loops were too small to determine the vacancy or interstitial nature of the dislocation loops directly. Although it is an indirect method, post-irradiation annealing was used to infer the loop nature. Most of dislocation loops were stable after the annealing at 400degC for 30 min. This result suggests that their nature is interstitial. (author)

  11. Stress corrosion cracking studies of reactor pressure vessel steels. Final report

    International Nuclear Information System (INIS)

    Van Der Sluys, W.A.

    1996-10-01

    The objective of this project was to perform a critical review of the information available in open literature on stress corrosion cracking of reactor pressure vessel materials in simulated light-water-reactor (LWR) conditions, develop a test procedure for conducting stress corrosion crack growth experiments in simulated LWR environments, and conduct a test program in an effort to duplicate some of the data available from the literature. The authors concluded that stress corrosion crack growth has been observed in pressure vessel steels under laboratory test conditions. The composition of the water in most cases where growth was observed is outside of the composition specified for operating conditions. Crack growth was observed in the experiments performed in this program, and it was intermittent. The cracking would start and stop for no apparent reason. In most instances, it would not restart without the change of some external variable. In a few instances, it restarted on its own. Crack growth rates as high as 3.6 x 10 -9 m/sec were observed in pressure vessel steels in high-purity water with 8 ppm oxygen. These high crack growth rates were observed for extremely short bursts in crack extension. They could not be sustained for crack growth extensions greater than a few tenths of a millimeter. From the results of this project it appears highly unlikely that stress corrosion cracking will be observed in operating nuclear plants where the coolant composition is maintained within water chemistry guidelines. However, more work is needed to better define the contaminations that cause crack growth. The crack growth rates are so high and the threshold values for crack nucleation are so low that the conditions causing them need to be well defined and avoided

  12. Application of electron beam welding to large size pressure vessels made of thick low alloy steel

    International Nuclear Information System (INIS)

    Kuri, S.; Yamamoto, M.; Aoki, S.; Kimura, M.; Nayama, M.; Takano, G.

    1993-01-01

    The authors describe the results of studies for application of the electron beam welding to the large size pressure vessels made of thick low alloy steel (ASME A533 Gr.B cl.2 and A533 Gr.A cl.1). Two major problems for applying the EBW, the poor toughness of weld metal and the equipment to weld huge pressure vessels are focused on. For the first problem, the effects of Ni content of weld metal, welding conditions and post weld heat treatment are investigated. For the second problem, an applicability of the local vacuum EBW to a large size pressure vessel made of thick plate is qualified by the construction of a 120 mm thick, 2350 mm outside diameter cylindrical model. The model was electron beam welded using local vacuum chamber and the performance of the weld joint is investigated. Based on these results, the electron beam welding has been applied to the production of a steam generator for a PWR. (author). 3 refs., 10 figs., 4 tabs

  13. Fatigue of non-welded pressure vessels made of high strength steel

    International Nuclear Information System (INIS)

    Rauscher, F.

    2003-01-01

    When using high strength steels for pressure vessels, cyclic fatigue requirements may become decisive for the design. Within a European research project, two typical non-welded types of vessels--gas cylinders as used for gas transportation and hydraulic accumulators with screwed in ends--were investigated. The results of the fatigue analyses and of the testing of these vessels are described here. Special attention is drawn to the evaluation of the stresses in the threads used for threaded in flat ends and rings, because the usual formulae for bolted connections cannot be used. In the case of sharp notches and of threads, the experiments showed that the fatigue calculation gave conservative results. The unexpected failure of the gas cylinders in the cylindrical part and at the onset of the end showed that the fatigue analyses according to prEN13445-3 clause 18 is non-conservative for these surfaces without mechanical preparation, and need special consideration. Based on the investigations, a stress concentration factor for small fabrication notches and a new surface finish factor is proposed

  14. Fatigue of non-welded pressure vessels made of high strength steel

    Energy Technology Data Exchange (ETDEWEB)

    Rauscher, F

    2003-03-01

    When using high strength steels for pressure vessels, cyclic fatigue requirements may become decisive for the design. Within a European research project, two typical non-welded types of vessels--gas cylinders as used for gas transportation and hydraulic accumulators with screwed in ends--were investigated. The results of the fatigue analyses and of the testing of these vessels are described here. Special attention is drawn to the evaluation of the stresses in the threads used for threaded in flat ends and rings, because the usual formulae for bolted connections cannot be used. In the case of sharp notches and of threads, the experiments showed that the fatigue calculation gave conservative results. The unexpected failure of the gas cylinders in the cylindrical part and at the onset of the end showed that the fatigue analyses according to prEN13445-3 clause 18 is non-conservative for these surfaces without mechanical preparation, and need special consideration. Based on the investigations, a stress concentration factor for small fabrication notches and a new surface finish factor is proposed.

  15. Stress corrosion cracking of nuclear reactor pressure vessel and piping steels

    International Nuclear Information System (INIS)

    Speidel, M.O.; Magdowski, R.M.

    1988-01-01

    This paper presents an extensive investigation of stress corrosion cracking of nuclear reactor pressure vessel and piping steels exposed to hot water. Experimental fracture mechanics results are compared with data from the literature and other laboratories. Thus a comprehensive overview of the present knowledge concerning stress corrosion crack growth rates is provided. Several sets of data confirm that 'fast' stress corrosion cracks with growth rates between 10 -8 and 10 -7 m/s and threshold stress intensities around 20 MN m -3/2 can occur under certain conditions. However, it appears possible that specific environmental, mechanical and metallurgical conditions which may prevail in reactors can result in significantly lower stress corrosion crack growth rates. The presently known stress corrosion crack growth rate versus stress intensity curves are discussed with emphasis on their usefulness in establishing safety margins against stress corrosion cracking of components in service. Further substantial research efforts would be helpful to provide a data base which permits well founded predictions as to how stress corrosion cracking in pressure vessels and piping can be reliably excluded or tolerated. It is emphasized, however, that the nucleation of stress corrosion cracks (as opposed to their growth) is difficult and may contribute substantially to the stress corrosion free service behaviour of the overwhelming majority of pressure vessels and pipes. (author)

  16. 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.)

  17. Thermal annealing of an embrittled reactor pressure vessel

    International Nuclear Information System (INIS)

    Mager, T.R.; Dragunov, Y.G.; Leitz, C.

    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 11 deals with thermal annealing of an embrittled reactor pressure vessel. Anneal procedures for vessels from both the US and the former USSR are mentioned schematically, wet anneals at lower temperature and dry anneals above RPV design temperatures are investigated. It is shown that heat treatment is a means of recovering mechanical properties which were degraded by neutron radiation exposure, thus assuring reactor pressure vessel compliance with regulatory requirements

  18. FP7 project LONGLIFE: Treatment of long-term irradiation embrittlement effects in RPV safety assessment

    International Nuclear Information System (INIS)

    May, J.; Hein, H.; Altstadt, E.; Bergner, F.; Viehrig, H.W.; Ulbricht, A.; Chaouadi, R.; Radiguet, B.; Cammelli, S.; Huang, H.; Wilford, K.

    2012-01-01

    The increasing age of European Nuclear Power Plants (NPPs) and envisaged extensions of plant lifetimes from 40 up to 80 years require an improved understanding of ageing phenomena of RPV components. The Network of Excellence NULIFE (Nuclear Plant Life Prediction) has been established to advance the safe and economic long-term operation (LTO) of NPPs by facilitating increased co-operation for applied R and D amongst members of the European nuclear community. The accurate prediction and management of RPV neutron irradiation embrittlement connected with long-term operation is an important aspect of this co-operation. Phenomena that might become important at high neutron fluences (such as flux effects and late blooming effects) have to be considered adequately in safety assessments. However, the surveillance database for prolonged irradiation times and low neutron fluxes is sparse. Consequently, there are significant uncertainties in the treatment of long-term irradiation effects. Therefore, the project LONGLIFE (Treatment of long-term irradiation embrittlement effects in RPV safety assessment) was initiated under the Euratom 7th Framework Programme of the European Commission as an umbrella project of NULIFE. LONGLIFE aims at 1) improved understanding of long-term irradiation phenomena that might compromise RPV integrity, and thereby the LTO of European NPPs, and 2) assessment of the adequacy of current prediction tools, codes, standards and surveillance guidelines for supporting long-term RPV operation. The scope of the work comprises the analysis of LTO boundary conditions; microstructural investigations and supplementary mechanical tests on RPV steels, including RPV steels from decommissioned plants; training activities; and elaboration of recommendations for RPV materials assessment and embrittlement surveillance under LTO conditions. A key part of the technical work is the selection of relevant materials for examination, e.g. which contain different weld and base

  19. Prediction of the Effects of Radiation FOr Reactor pressure vessel and in-core Materials using multi-scale modeling - 60 years foreseen plant lifetime (PERFORM-60 project)

    International Nuclear Information System (INIS)

    Al Mazouzi, A.; Bugat, S.; Leclercq, S.; Massoud, J.-P.; Moinereau, D.; Lidbury, D.; Van Dyck, S.; Marini, B.; Alamo, Ana

    2010-01-01

    The PERFECT project of the EURATOM framework program (FP6) is a first step through the development of a simulation platform that contains several advanced numerical tools aiming at the prediction of irradiation damage in both the reactor pressure vessel (RPV) and its internals using state-of-the-art knowledge. These tools allow simulation of irradiation effects on the microstructure and the constitutive behavior of the RPV low alloy steels, as well as their fracture mechanics properties. For the reactor internals, the first partial models were established, describing radiation damage to the microstructure and providing a first description of the stress corrosion behaviour of austenitic steels in primary environment, without physical linking of the radiation and corrosion effects. Thus, relying on the existing PERFECT Roadmap, the FP7 Collaborative Project PERFORM 60 has mainly for objective to develop similar tools that would allow the simulation of the combined effects of irradiation and corrosion on internals, in addition to a further improvement of the existing ones on RPV made of bainitic steels. From the managerial view point, PERFORM 60 is based on two technical sub-projects, namely (i) RPV and (ii) Internals. In addition, a Users' Group and a training scheme have been adopted in order to allow representatives of constructors, utilities, research organizations... from Europe, USA and Japan to participate actively in the process of appraising the limits and potentialities of the developed tools as well as their validation against qualified experimental data

  20. Study on structural failure of RPV with geometric discontinuity under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Mao, J.F., E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Zhu, J.W. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Department of Mechanical and Electrical engineering, Huzhou Vocational & Technical College Huzhou, Zhejiang 313000 (China); Bao, S.Y., E-mail: bsy@zjut.edu.cn [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China); Luo, L.J. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Gao, Z.L. [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Re-manufacturing, Ministry of Education (China)

    2016-10-15

    Highlights: • The RPV failure is investigated in depth under severe accident. • The creep and plastic damage are the major contributor to RPV failure. • A elastic core is found at the midpoint of the highly-eroded region. • Weakest location has some ‘accommodating’ quality to prevent ductile tearing. • The internal pressure is critical for the determination of structural failure. - Abstract: A severe accident management strategy known as ‘in-vessel retention (IVR)’ is widely adopted in most of advanced nuclear power plants. The IVR mitigation is assumed to be able to arrest the degraded melting core and maintain the structural integrity of reactor pressure vessel (RPV) within a prescribed period of time. This traditional concept of IVR without consideration of internal pressure effect wasn’t challenged until the occurrence of Fukushima accident on 2011, which showed that the structural behavior had not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still existed inside the RPV. Accordingly, the paper tries to address the related issue on whether lower head (LH) integrity can be maintained, when the LH is subjected to the thermal-mechanical loads created during such a severe accident. Because of the presence of the high temperature melt (∼1300 °C) on the inside of RPV, some local material is melted down to create a unique RPV with geometric discontinuity, while the outside of RPV submerged in cavity water will remain in nucleate boiling (at ∼150 °C). Therefore, the failure mechanisms of RPV can span a wide range of structural behaviors, such as melt-through, creep damage, plastic yielding as well as thermal expansion. Through meticulous investigation, it is found that the RPV failure is mainly caused by creep and plasticity, especially for the inside of highly-eroded region. The elastic core (or layer) is found to exist in the proximity of mid-section of the highly-eroded wall. However, the elastic core is squeezed into

  1. Soft RPV through the baryon portal

    International Nuclear Information System (INIS)

    Krnjaic, Gordan; Tsai, Yuhsin

    2014-01-01

    Supersymmetric (SUSY) models with R-parity generically predict sparticle decays with invisible neutralinos, which yield distinctive missing energy events at colliders. Since most LHC searches are designed with this expectation, the putative bounds on sparticle masses become considerably weaker if R-parity is violated so that squarks and gluinos decay to jets with large QCD backgrounds. Here we introduce a scenario in which baryonic R-parity violation (RPV) arises effectively from soft SUSY breaking interactions, but leptonic RPV remains accidentally forbidden to evade constraints from proton decay and FCNCs. The model features a global R-symmetry that initially forbids RPV interactions, a hidden R-breaking sector, and a heavy mediator that communicates this breaking to the visible sector. After R-symmetry breaking, the mediator is integrated out and an effective RPV A-term arises at tree level; RPV couplings between quarks and squarks arise only at loop level and receive additional suppression. Although this mediator must be heavy compared to soft masses, the model introduces no new hierarchy since viable RPV can arise when the mediator mass is near the SUSY breaking scale. In generic regions of parameter space, a light thermally-produced gravitino is stable and can be a viable dark matter candidate

  2. Effects of ATR-2 Irradiation to High Fluence on Nine RPV Surveillance Materials

    Energy Technology Data Exchange (ETDEWEB)

    Nanstad, Randy K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Odette, George R. [Univ. of California, Santa Barbara, CA (United States); Almirall, Nathan [Univ. of California, Santa Barbara, CA (United States); Robertson, Janet [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Server, W. L. [ATI Consulting, Pinehurst, NC (United States); Yamamoto, T. [Univ. of California, Santa Barbara, CA (United States); Wells, Peter [Univ. of California, Santa Barbara, CA (United States)

    2017-05-01

    The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely degraded, with the degree of toughness loss dependent on the radiation sensitivity of the materials. The available embrittlement predictive models and our present understanding of radiation damage are not fully quantitative, and do not treat all potentially significant variables and issues, particularly considering extension of operation to 80y.

  3. The analysis of RPV fast neutron flux calculation for PWR with three-dimensional SN method

    International Nuclear Information System (INIS)

    Yang Shouhai; Chen Yixue; Wang Weijin; Shi Shengchun; Lu Daogang

    2011-01-01

    Discrete ordinates (S N ) method is one of the most widely used method for reactor pressure vessel (RPV) design. As the fast development of computer CPU speed and memory capacity and consummation of three-dimensional discrete-ordinates method, it is mature for 3-D S N method to be used to engineering design for nuclear facilities. This work was done specifically for PWR model, with the results of 3-D core neutron transport calculation by 3-D core calculation, 3-D RPV fast neutron flux distribution obtain by 3-D S N method were compared with gained by 1-D and 2-D S N method and the 3-D Monte Carlo (MC) method. In this paper, the application of three-dimensional S N method in calculating RPV fast neutron flux distribution for pressurized water reactor (PWR) is presented and discussed. (authors)

  4. Heavy-section steel irradiation program: Embrittlement issues

    International Nuclear Information System (INIS)

    Corwin, W.R.

    1991-01-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. The RPV is one of only two major safety- related components of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Effects of specimen size, material chemistry, product form and microstructure, irradiation fluence, flux, temperature and spectrum, and postirradiation annealing are being examined on a wide range of fracture properties including fracture toughness crack arrest toughness ductile tearing resistance Charpy V-notch impact energy, dropweight nil-ductility temperature and tensile properties. Models based on observations of radiation-induced microstructural changes using the field on microprobe and the high resolution transmission electron microscopy provide improved bases for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs

  5. Heavy-Section Steel Irradiation Program: Embrittlement issues

    International Nuclear Information System (INIS)

    Corwin, W.R.

    1991-01-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents and the potential for major contamination releases. It is imperative to understand and predict the capabilities and limitations of its integrity. It is particularly vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage it is virtually impossible to postulate a realistic scenario which would result in RPV failure. The Heavy-Section Steel Irradiation (HSSI) Program has been established by the US Nuclear Regulatory Commission (USNRC) to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure vessel steels as they relate to light-water reactor pressure-vessel integrity. Results from HSSI studies provide information needed to aid in resolving major regulatory issues facing the USNRC which involve RPV irradiation embrittlement such as pressurized-thermal shock, operating pressure-temperature limits, low-temperature overpressurization, and the specialized problems associated with low upper-shelf (LUS) welds. Taken together the results of these studies also provide guidance and bases for evaluating both the aging behavior and the potential for plant life extension of light-water RPVs. The principal materials examined within the HSSI program are high-copper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. Embrittlement modeling studies have shown that the time or dose required for the point defect concentrations, which ultimately contribute to irradiation embrittlement, to reach their steady state values can be comparable to the component lifetime or to the duration of an irradiation experiment

  6. Characteristics of Modified 9Cr-1Mo Steel for Reactor Pressure Vessel of Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Ryu, W. S.; Han, Chang Hee; Yoon, J. H.; Chang, Jong Hwa

    2004-11-15

    Many researches and developments have been progressed for the construction of VHTR by 2020 in Korea. Modified 9Cr-1Mo steel has been receiving attention for the application to the reactor pressure vessel material of VHTR. We collected and analyzed the research data for modified 9Cr-1Mo steel in order to understand the characteristics of modified 9Cr-1Mo steel. The modified 9Cr-1Mo steel is a modified alloy system similar to conventional 9Cr-1Mo grade ferritic steel. Modifications include additions of vanadium, niobium, and nitrogen, as well as lower carbon content. In this report, we summarized the change of microstructure and mechanical properties after tempering, thermal aging, and irradiation. Modified 9Cr-1Mo steel has high strength and thermal conductivity, low thermal expansion, and good resistance to corrosion. But the irradiation embrittlement behavior of modified 9Cr-1Mo steel should be evaluated and the evaluation methodology also should be developed. At the same time, the characteristics of weldment which is the weak part in pressure vessel should be evaluated.

  7. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  8. Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

    International Nuclear Information System (INIS)

    Komine, Kuniaki; Konno, Mutsuo

    1999-01-01

    NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

  9. Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

    International Nuclear Information System (INIS)

    Burke, M.G.; Freyer, P.D.; Mager, T.R.

    1993-01-01

    In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ''precipitation-type'' and a ''damage-type'' component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs

  10. Round Robin Posttest analysis of a 1/10-scale Steel Containment Vessel Model Test

    Energy Technology Data Exchange (ETDEWEB)

    Komine, Kuniaki [Nuclear Power Engineering Corp., Tokyo (Japan); Konno, Mutsuo

    1999-07-01

    NUPEC and U.S. Nuclear Regulatory Commission (USNRC) have been jointly sponsoring 'Structural Behavior Test' at Sandia National Laboratory (SNL) in Cooperative Containment Research Program'. As one of the test, a test of a mixed scaled SCV model with 1/10 in the geometry and 1/4 in the shell thickness. Round Robin analyses of a 1/10-scale Steel Containment Vessel (SCV) Model Test were carried out to obtain an adequate analytical method among seven organizations belonged to five countries in the world. As one of sponsor, Nuclear Power Engineering Corporation (NUPEC) filled the important role of a posttest analysis of SCV model. This paper describes NUPEC's analytical results in the round robin posttest analysis. (author)

  11. 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.)

  12. Long-term ageing effects in reactor pressure vessel steels investigated by positron annihilation spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    Butterling, Maik; Anwand, Wolfgang; Wagner, Andreas [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Radiation Physics, Dresden (Germany); Bergner, Frank; Ulbricht, Andreas; Wagner, Arne [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Ion Beam Physics and Materials Research, Dresden (Germany)

    2014-07-01

    Neutron irradiation of reactor pressure vessel steels leads to the formation of nano-sized defects which can deteriorate the material. An understanding of the microstructural evolution of the material is important for making reliable security assessments about possible future long-term operation of nuclear power plants. So-called late-blooming phases are formed after long-term irradiation and lead to considerable material ageing effects. Encouraging factors for the formation of these phases are a low Cu-content, moderate to high contents of Mn and Ni, low irradiation temperatures and different neutron fluxes. Positron annihilation lifetime spectroscopy which is ideally suited for the detection and characterization of these irradiation-induced defects was applied for different selected materials which fulfill these conditions in order to investigate the occurrence and behavior of these phases.

  13. Irradiation embrittlement of reactor pressure vessel steels: Considerations for thermal annealing

    Energy Technology Data Exchange (ETDEWEB)

    Burke, M G; Freyer, P D; Mager, T R

    1994-12-31

    In this paper, an overview of the irradiation embrittlement phenomenon is presented from a structure-properties viewpoint. Effects of irradiation conditions on embrittlement are first reviewed: irradiation temperature, fluence, flux, and steel or alloy composition. Then, the techniques for identifying/characterizing the irradiation-induced microstructural features are described: TEM/STEM (electron microscopy), small angle neutron scattering, atom probe field-ion microscopy, positron annihilation lifetime spectroscopy. Mechanisms of hardening and embrittlement generally consist of a ``precipitation-type`` and a ``damage-type`` component and the potential of annealing treatments for restoring the most of the original pressure vessel material toughness is examined; its conditions and mechanisms involved are discussed. Feasibility and economic evaluation of annealing costs is also carried out. 90 refs., 4 figs.

  14. Microstructural parameters and yielding in a quenched and tempered Cr-Mo-V pressure vessel steel

    International Nuclear Information System (INIS)

    Toerroenen, Kari

    1979-01-01

    In this work the plastic deformation behaviour of a Cr-Mo-V pressure vessel steel is studied at ambient and low temperatures. To produce a wide range of microstructures, different austenitizing, quenching and tempering treatments are performed. The microstructures, including grain and dislocation structures as well as carbides, are evaluated. A qualitative model is proposed for the martensitic and bainitic transformations explaining the morphology and crystallography of the transformation products. Based on microstructural observations of undeformed and deformed materials, as well as the tensile test results, the role of various obstacles to dislocation motion in plastic deformation is evaluated. Finally the strength increment, its temperature dependence and the effect due to combinations of various obstacles are analyzed. The results are intended to serve as basis for further fracture behaviour analyses. (author)

  15. 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)

  16. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Lee, A.D.

    1997-01-01

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  17. The failure behavior of duplex 316 L steel-TA6V titanium alloy spherical pressure vessels

    International Nuclear Information System (INIS)

    Miannay, D.

    1980-05-01

    The purpose of this paper is to compare the experimental residual stresses of spherical vessels made of TA6V alloy which exhibits plasticity before failure in toughness testing and cracked with several configurations, with stresses estimated according to the afore mentioned theories. An internal austenitic 316 L steel is used to prevent 'leak before break' [fr

  18. Interaction between molten corium UO2+x-ZrO2-FeOy and VVER vessel steel

    International Nuclear Information System (INIS)

    Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A.; Gusarov, V. V.; Almiashev, V. I.; Lopukh, D. B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

    2010-01-01

    In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO 2+x -ZrO 2 -FeO y melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction Zone. The available experimental data have been used to develop a correlation for the corrosion rate as a function of temperature and heat flux. (authors)

  19. Stereofractographic investigation of static start and dynamic jump of a fatigue crack in pressure vessel steel

    International Nuclear Information System (INIS)

    Stepanenko, V.A.; Shtukaturova, A.S.; Yasnij, P.V.; AN Ukrainskoj SSR, Kiev. Inst. Problem Prochnosti)

    1983-01-01

    Results of investigaytion have been discussed ipto the effect of certain temperature-force factors on regularities in the formation of stretch zones during the crack initiation static and transition zones at crack jumps in the process of cyclic loading. The 15Kh2NMFA pressure vessel steel has been investigated. The steel fracture toughnesKub(Ic) has been determined testing s the specimens for excentric stretching or a bending through an angle. It has been shown that transition zones in a front of fatique cracks at the jump beginning and end are formed through the shift mechanism owing to the material separation along the maximum failure zone contour, i.e. along the plastic zone contour in a crack vertex. This is the mait difference of regularities in the formation of the transition zones during fatique crack jumps from stretching zones formed through the break-away mechanism of crack vertex bluntness during its static move. It is noted that a final conclusion on the mechanism of transition zone formation during fartique crack jumps allows one to perform systematic investigation into the plastic zone configuration in a fatique crack verteX and stereofractographic measurement of two identically conjugate jump surfaces on opposite fractures of the same samples

  20. Micromechanisms of ductile stable crack growth in nuclear pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Belcher, W.P.A.; Druce, S.G.

    1981-10-01

    The objective of this work was to investigate the relationship between the micromechanisms of ductile crack growth, the microstructural constituent phases present in nuclear pressure vessel steel, and the observed fracture behavior as determined by impact and fracture mechanics tests. Results from a microstructural and mechanical property comparison of an A508 Class 3 pressurized water reactor nozzle forging cutout and a 150-mm-thick A533B Class 1 plate are reported. The variation of upper-shelf toughness between the two steels and its orientation sensitivity are discussed on the basis of inclusion and precipitate distributions. Inclusion clusters in A533B, deformed to elongated disks in the rolling plane, have a profound effect on short transverse fracture properties. Data derived using the multi-specimen J-integral method to characterize the initiation of ductile crack extension and resistance to stable crack growth are compared with equivalent Charpy results. Results of the J /SUB R/ -curve analyses indicate (1) that the A533B short transverse crack growth resistance is approximately half that observed from transverse and longitudinal specimen orientations, and (2) that the A508 initiation toughness and resistance to stable crack growth are insensitive to position through the forging wall, and are higher than exhibited by A533B at any orientation in the midthickness position.

  1. Microstructure and mechanical characteristics of a laser welded joint in SA508 nuclear pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wei, E-mail: wei.guo-2@manchester.ac.uk [Laser Processing Research Centre, School of Mechanical, Aerospace and Civil Engineering, University of Manchester, Sackville Street, Manchester, M13 9 PL (United Kingdom); Dong, Shiyun [Laser Processing Research Centre, School of Mechanical, Aerospace and Civil Engineering, University of Manchester, Sackville Street, Manchester, M13 9 PL (United Kingdom); Institute of Laser Engineering, Beijing University of Technology, Beijing 100124 (China); Guo, Wei; Francis, John A.; Li, Lin [Laser Processing Research Centre, School of Mechanical, Aerospace and Civil Engineering, University of Manchester, Sackville Street, Manchester, M13 9 PL (United Kingdom)

    2015-02-11

    SA508 steels are typically used in civil nuclear reactors for critical components such as the reactor pressure vessel. Nuclear components are commonly joined using arc welding processes, but with design lives for prospective new build projects exceeding 60 years, new welding technologies are being sought. In this exploratory study, for the first time, autogenous laser welding was carried out on 6 mm thick SA508 Cl.3 steel sheets using a 16 kW fiber laser system operating at a power of 4 kW. The microstructure and mechanical properties (including microhardness, tensile strength, elongation, and Charpy impact toughness) were characterized and the microstructures were compared with those produced through arc welding. A three-dimensional transient model based on a moving volumetric heat source model was also developed to simulate the laser welding thermal cycles in order to estimate the cooling rates included by the process. Preliminary results suggest that the laser welding process can produce welds that are free of macroscopic defects, while the strength and toughness of the laser welded joint in this study matched the values that were obtained for the parent material in the as-welded condition.

  2. 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)

  3. Fracture toughness behavior and its analysis on nuclear pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Iwadate, Tadao; Tanaka, Yasuhiko; Ono, Shin-ichi; Tsukada, Hisashi [Japan Steel Works Ltd., Muroran, Hokkaido. Muroran Plant

    1983-02-01

    A drop weight J sub(Id) testing machine has been developed successfully, by which the multiple specimen J resistance curve test technique can be applied to measure the fracture toughness. In this study, the use of a small size round compact tension (RCT) specimen for measuring the fracture toughness J sub(Ic) or J sub(Id) of the nuclear pressure vessel steels is recommended and confirmed for the surveillance tests. The static and dynamic fracture toughness of ASTM A508 C 1.2, A508 C 1.3 and A533 Gr.B C 1.1 steels in the wide range of temperature including the upper shelf have been measured and their behavior has been analysed. The fracture toughness behavior under various strain rates and in a wide temperature range can be explained by the behavior of stretched zone formation preceding the crack initiation. The scatter of K sub(J) values in the transition range is caused by the amount of crack extension contained in the specimens. In this paper, the method to obtain the fracture toughness equivalent to the K sub(Ic) from the K sub(J) value is also presented.

  4. Effect of Macrosegregation on the Microstructure and Mechanical Properties of a Pressure-Vessel Steel

    Science.gov (United States)

    Yan, Guanghua; Han, Lizhan; Li, Chuanwei; Luo, Xiaomeng; Gu, Jianfeng

    2017-07-01

    Macrosegregation refers to the chemical segregation, which occurs quite commonly in the large forgings such as nuclear reactor pressure vessel. This work assesses the effect of macrosegregation and homogenization treatment on the mechanical properties of a pressure-vessel steel (SA508 Gr.3). It was found that the primary reason for the inhomogeneity of the microstructure was the segregation of Mn, Mo, and Ni. Martensite, and coarse upper bainite with M-A (martensite-austenite) islands have been obtained, respectively, in the positive and negative segregation zone during a simulated quenching process. During tempering, the carbon-rich M-A islands decomposed into a mixture of ferrite and numerous carbides which deteriorated the toughness of the material. The segregation has been substantially minimized by a homogenizing treatment. The results indicate that the material homogenized has a higher impact toughness than the material with segregation, due to the reduction in M-A island in the negative segregation zone. It can be concluded that the microstructure and mechanical properties have been improved remarkably by means of homogenization treatment.

  5. Acoustoelastic evaluation of welding and heat treatment stress relieving of pressure vessel steel for Angra 3

    Energy Technology Data Exchange (ETDEWEB)

    Moraes, Bruno C. de, E-mail: bruno.cesar@nuclep.gov.br [Nuclebras Equipamentos Pesados S.A (NUCLEP), Itaguai, RJ (Brazil); Bittencourt, Marcelo de S.Q., E-mail: bruno.cesar@nuclep.gov.br, E-mail: bittenc@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Currently the knowledge of non-destructive techniques allows to evaluate the stresses on components and mechanical structures, aiming at physical security, preservation of the environment and avoid financial losses associated with the construction and operation of industrial plants. The search for new techniques, especially applied in the nuclear industry to assess status more accurately, voltage safety and to ensure structural integrity, for example, core components of the primary circuit, such as the reactor pressure vessel and steam generator has become of great importance within the community of non-destructive testing .This paper aims to contribute to the non-destructive technique development in order to ensure the structural integrity of nuclear components. One acoustoelastic evaluation of steel 20 MnMoNi 55, used in pressure vessels of nuclear power plants were performed. The acoustic birefringence technique was use to evaluate the acoustoelastic behavior of the test material in the as received condition, after welding and after the stress relief heat treatment. The constant acoustoelastic material was obtained by an uniaxial loading test. It was found a slight anisotropy in the material as received. After welding, a marked variation of acoustic birefringence in the region near the weld bead was observed. The heat treatment indicated a new change of acoustic birefringence. Obtaining the acoustoelastic constant allowed the evaluation of stress in the different conditions of the weld and treated material. (author)

  6. Acoustoelastic evaluation of welding and heat treatment stress relieving of pressure vessel steel for Angra 3

    International Nuclear Information System (INIS)

    Moraes, Bruno C. de; Bittencourt, Marcelo de S.Q.

    2015-01-01

    Currently the knowledge of non-destructive techniques allows to evaluate the stresses on components and mechanical structures, aiming at physical security, preservation of the environment and avoid financial losses associated with the construction and operation of industrial plants. The search for new techniques, especially applied in the nuclear industry to assess status more accurately, voltage safety and to ensure structural integrity, for example, core components of the primary circuit, such as the reactor pressure vessel and steam generator has become of great importance within the community of non-destructive testing .This paper aims to contribute to the non-destructive technique development in order to ensure the structural integrity of nuclear components. One acoustoelastic evaluation of steel 20 MnMoNi 55, used in pressure vessels of nuclear power plants were performed. The acoustic birefringence technique was use to evaluate the acoustoelastic behavior of the test material in the as received condition, after welding and after the stress relief heat treatment. The constant acoustoelastic material was obtained by an uniaxial loading test. It was found a slight anisotropy in the material as received. After welding, a marked variation of acoustic birefringence in the region near the weld bead was observed. The heat treatment indicated a new change of acoustic birefringence. Obtaining the acoustoelastic constant allowed the evaluation of stress in the different conditions of the weld and treated material. (author)

  7. The non-destructive examination of reactor pressure vessel steels by positron annihilation

    International Nuclear Information System (INIS)

    Highton, J.P.

    1983-01-01

    The rapid radiation hardening of copper bearing reactor pressure vessel steels has been linked with microvoids that are associated with copper based complexes in the metal lattice. These microvoids are active in the sense that their size appears to be related to the temperature of irradiation, which thus determines their influence on dislocation mobility. These sites appear to grow by vacancy condensation which causes a reduction in the local lattice energy. Thus prolonged exposure to PWR temperatures, even in the absence of a neutron flux, may also cause embrittlement. It has been found that these sites, which represent a local negative charge, act as traps to positrons. The size of each site dictates its positron trapping potential. As the trapping potential increases so too does the probability that the positrons will annihilate with low momentum conduction electrons. The momentum of the annihilating electrons will determine the degree of Doppler broadening of the 511 keV annihilation gamma peak. Thus careful analysis of this peak can yield useful information on the degree of embrittlement caused by these active defect complexes. In this way positron annihilation offers a powerful non-destructive alternative to current methods of assessing the integrity of nuclear reactor pressure vessels. (author)

  8. Irradiation induced tensile property change of SA 508 Cl.3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se-Hwan; Hong, Jun-Hwa; Kuk, Il-Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the unirradiated and irradiated microstructure. Microvickers hardness, indentation, and miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were 2 irradiated to a neutron fluence of 2.7x10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg. C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Band-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural. state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation(VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by

  9. Irradiation induced tensile property change of SA 508 Cl. 3 reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Chi, Se Hwan; Hong, Jun Hwa; Kuk, Il Hiun

    1998-01-01

    Irradiation induced tensile property change of four kinds of reactor pressure vessel steels manufactured by different steel refining process was compared based on the differences in the miniature tensile specimen tests were conducted for mechanical property measurement and optical microscope (OM) and transmission electron microscope (TEM) were used for microstructural characterization. Specimens were irradiated to a neutron fluence of 2.7 x 10 19 n/cm 2 (E ≥ 1 MeV) at 288 deg C. Investigation on the unirradiated microstructures showed largely a same microstructure in that tempered acicular bainite and ferrite with bainitic phase prevailing in the unirradiated condition. Ban-shaped segregations were also clearly observed except a kind of materials. A large difference in the unirradiated microstructure appeared in the grain size and carbide microstructure. Of carbide microstructures, noticeable differences were observed in the size and distribution of cementite, and bainitic lath microstructures. No noticeable changes were observed in the optical and thin film TEM microstructures after irradiation. Complicated microstructural state of heat treated bainitic low alloy microstructure prevents easy quantification of microstructural changes due to irradiation. Apparent differences, however, were observed in the results of mechanical testing. Results of tensile testing and hardness measurement show that a steel refined by vacuum carbon deoxidation (VCD) method exhibits the highest radiation hardening behavior. Some of mechanical testing results on irradiated materials were possible to understand based on the initial microstructure, but further investigations using a wide array of sophisticated tools (for example, SANS, APFIM) are required to understand and characterize irradiation induced defects that are responsible for irradiation hardening behavior but are not revealed by conventional TEM. (author)

  10. Fracture toughness determination of the pressure vessel steel A508 Cl 2 between 100 and 350 degree C

    International Nuclear Information System (INIS)

    Rao, S.

    1980-09-01

    The fracture toughness of the pressure vessel steel A508 was determined in the temperature range 100 - 350 degree C. The J-integral method with crack growth resistance curves, the so-called R-curves, was used. The results show that the steel does not have an 'upper-shelf' and the fracture toughness, K sub (JC), decreases with increasing temperature to a minimum around 300 degree C and an increase above it. These results are compared to those obtained previously on an other pressure vessel steel A533B which has essentially the same temperature dependence. The results were also analysed using the Tearing modulus, T. The conclusion iw that the crack growth resistance and the crack initiation resistance (K sub (JC)) show a significant decrease around the operating temperatures as compared to 100 degree C. (author)

  11. Structural mechanisms of the flux effect for VVER-1000 reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Gurovich, B.; Kuleshova, E.; Fedotova, S.; Maltsev, D.; Zabusov, O.; Frolov, A.; Erak, D.; Zhurko, D.

    2015-01-01

    To justify the lifetime extension of VVER-1000 reactor pressure vessels (RPV) up to 60 years and more it is necessary to expand the existing surveillance samples database to beyond design fluence by means of accelerated irradiation in a research reactor. Herewith since the changes in mechanical properties of materials under irradiation are due to occurring structural changes, correct analysis of the data obtained at accelerated irradiation of VVER-1000 RPV materials requires a clear understanding of the structural mechanisms that are responsible for the flux effect in VVER-1000 RPV steels. Two mechanisms are responsible for radiation embrittlement of VVER-1000 RPV steels: the hardening one (radiation hardening due to formation of radiation-induced Ni-based precipitates and radiation defects) and non-hardening one (due to formation of impurities segregations at grain boundaries - reversible temper brittleness). In this context for an adequate interpretation of the mechanical tests results when justifying the lifetime extension of existing units a complex of comparative structural studies (TEM, SEM and AES) of VVER-1000 RPV materials irradiated in different conditions (in research reactor IR-8 and within surveillance samples) was performed. It is shown that the flux effect is observed for materials with high nickel content (weld metals with Ni content > 1.35%) and it is mostly due to the contribution of non-hardening mechanism of radiation embrittlement (the difference in the accumulation kinetics of grain boundary phosphorus segregation) and somewhat contribution of the hardening mechanism (the difference in density of radiation-induced precipitates). Therefore when analyzing the results obtained from the accelerated irradiation of VVER-1000 WM the correction for the flux effect should be made. (authors)

  12. The influence of fire exposure on austenitic stainless steel for pressure vessel fitness-for-service assessment: Experimental research

    Science.gov (United States)

    Li, Bo; Shu, Wenhua; Zuo, Yantian

    2017-04-01

    The austenitic stainless steels are widely applied to pressure vessel manufacturing. The fire accident risk exists in almost all the industrial chemical plants. It is necessary to make safety evaluation on the chemical equipment including pressure vessels after fire. Therefore, the present research was conducted on the influences of fire exposure testing under different thermal conditions on the mechanical performance evolution of S30408 austenitic stainless steel for pressure vessel equipment. The metallurgical analysis described typical appearances in micro-structure observed in the material suffered by fire exposure. Moreover, the quantitative degradation of mechanical properties was investigated. The material thermal degradation mechanism and fitness-for-service assessment process of fire damage were further discussed.

  13. Practical implications for RPV irradiation surveillance under long term operation based on latest research results

    International Nuclear Information System (INIS)

    Hein, H.; Keim, E.; Barthelmes, J.; Schnabel, H.

    2015-01-01

    The international programs CARISMA, CARINA and LONGLIFE belong to the research programs which have been performed during the last 10 years to study the irradiation behavior of RPV steels under long term operation of more than 60 years. Some characteristic but different irradiated RPV steels used in Pressurized Water Reactors have been extensively investigated in each of those three programs. Whereas the CARISMA and CARINA programs were mainly focused on material testing to study the irradiation-induced change of material properties in terms of fracture toughness, the main objective of LONGLIFE was to investigate the change of microstructure with various analysis techniques and to understand the mechanisms behind. In this way it was possible to get a comprehensive material characterization in terms of macro-physical properties and micro-structural features for a number of RPV steels which have been studied at different irradiation levels up to 8*10 19 cm -2 (E > 1 MeV). The essential macro-physical and micro-structural results are summarized, in particular regarding the impact of copper and nickel, and the neutron flux on the irradiation behavior and with respect to possible late irradiation effects under long term operation. Moreover, the change of material properties is linked with embrittlement mechanisms such as formation of element specific precipitations, segregations, and matrix defects. Well-known trend curves are also applied to the measured T 41 and T 0 data in order to assess their appropriateness for long term operation. Based on the comprehensive available data base, practical implications for RPV irradiation surveillance programs under long term operation are highlighted with respect to issues like material specific application of reference temperature concepts, data scattering, prediction of high fluence behavior and how to cope with possible late irradiation effects. Finally, best practices for RPV irradiation surveillance programs are suggested from

  14. Manufacturing of ultra-large diameter 20 MnMoNi 5 5 steel forgings for reactor pressure vessels and their properties

    International Nuclear Information System (INIS)

    Onodera, S.; Kawaguchi, S.; Tsukada, H.; Moritani, H.; Suzuki, K.; Sato, I.

    1985-01-01

    As the structural material for RPV typical of increased dimensions, as set of ultra-large diameter steel forgings for a PHWR RPV is presented as outlined below. (1) Material designation: 20 MnMoNi 5 5 (similar to SA508, Cl.3) (2) Size of the forgings: flanges, 8.440 mm OD, a weight of 238 tons for shell flange; shells and torus, 7,920 mm OD, with large height, cover dome, 6,800 mm OD in chord and 460 mm thick; blank before formed to dome is ca. 8,000 mm OD. (3) Chemical composition: particular effort was made for minimizing the tramp elements as P, S, As, Sn, Sb, Cu. (4) Manufacturing, key points: steel making - combined refining and degassing in ladle; ingot making - largest size ingots, including 570 ton and 500 ton ingots; forging - special ''outside-the-press'' forging and forming techniques; heat treatment - prevention of H 2 flaking in normalizing and tempering and handling of the extra-large forgings at water quenching. (5) Metallurgical properties: sufficiently uniform carbon distributions in the forgings; a lowest possible content of hydrogen, non-metallic inclusions and oxygen. Mechanical properties: uniformity in tensile and toughness properties; flaws - only limited number of spots of UT indications under 2 mm EFG (EFS). (orig.)

  15. Heavy section steel technology program technical report No. 38. Fracture toughness characterization of HSST intermediate pressure vessel material

    International Nuclear Information System (INIS)

    Mager, T.R.; Yanichko, S.E.; Singer, L.R.

    1974-12-01

    The primary objective of the Heavy Section Steel Technology (HSST) Program is to develop pertinent fracture technology to demonstrate the structural reliability of present and contemplated water-cooled nuclear reactor pressure vessels. In order to demonstrate the ability to predict failure of large, heavy-walled pressure vessels under service type loading conditions, the fracture toughness properties of the vessel's materials must be characterized. The sampling procedure and test results are presented for vessel material supplied by the Oak Ridge National Laboratory that were used to characterize the fracture toughness of the HSST Intermediate Test Vessels. The metallurgical condition and heat treatment of the test material was representative of the vessel simulated service test condition. Test specimen locations and orientations were selected by the Oak Ridge National Laboratory and are representative of flaw orientations incorporated in the test vessels. The fracture toughness is documented for the materials from each of the eight HSST Intermediate Pressure Vessels tested to date. 7 references. (U.S.)

  16. Development of a supplemental surveillance program for reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Server, W.L.; Rosinski, S.T.

    1997-01-01

    The technical decision to thermally anneal a nuclear reactor pressure vessel (RPV) depends upon the level of embrittlement in the RPV steels, the amount of recovery of fracture toughness properties expected from the anneal, and the rate of re-embrittlement after the vessel is placed back into service. The recovery of Charpy impact toughness properties after annealing can be estimated initially by using a recovery model developed using experimental measurements of recovery (such as that developed by Eason et al. for U.S. vessel materials). However, actual validation measurements on plant-specific archived vessel materials (hopefully in the existing surveillance program) are needed; otherwise, irradiated surrogate materials, essentially the same as the RPV steels or bounding in expected behavior, must be utilized. The efficient use of any of these materials requires a supplemental surveillance program focused at both recovery and reirradiation embrittlement. Reconstituted Charpy specimens and new surveillance capsules will most likely be needed as part of this supplemental surveillance program. A new version of ASTM E 509 has recently been approved which provides guidance on thermal annealing in general and specifically for the development of an annealing supplemental surveillance program. The post-anneal re-embrittlement properties are crucial for continued plant operation, and the use of a re-embrittlement model, such as the lateral shift approach, may be overly conservative. This paper illustrates the new ASTM E 509 Standard Guide methodology for an annealing supplemental surveillance program. As an example, the proposed program for the Palisades RPV beltline steels is presented which covers the time from annealing to the end of operating license and beyond, if license renewal is pursued. The Palisades nuclear power plant RPV was planned to be annealed in 1998, but that plant is currently being re-evaluated. The proposed anneal was planned to be conducted at a

  17. Pressurized thermal shock in nuclear power plants: Good practices for assessment. Deterministic evaluation for the integrity of reactor pressure vessel

    International Nuclear Information System (INIS)

    2010-02-01

    Starting in the early 1970s, a series of coordinated research projects (CRPs) was sponsored by the IAEA focusing on the effects of neutron radiation on reactor pressure vessel (RPV) steels and RPV integrity. In conjunction with these CRPs, many consultants meetings, specialists meetings, and international conferences, dating back to the mid-1960s, were held. Individual studies on the basic phenomena of radiation hardening and embrittlement were also performed to better understand increases in tensile strength and shifts to higher temperatures for the integrity of the RPV. The overall objective of this CRP was to perform benchmark deterministic calculations of a typical pressurized thermal shock (PTS) regime, with the aim of comparing the effects of individual parameters on the final RPV integrity assessment, and then to recommend the best practices for their implementation in PTS procedures. At present, several different procedures and approaches are used for RPV integrity assessment for both WWER 440-230 reactors and pressurized water reactors (PWRs). These differences in procedures and approaches are based, in principle, on the different codes and rules used for design and manufacturing, and the different materials used for the various types of reactor, and the different levels of implementation of recent developments in fracture mechanics. Benchmark calculations were performed to improve user qualification and to reduce the user effect on the results of the analysis. This addressed generic PWR and WWER types of RPV, as well as sensitivity analyses. The complementary sensitivity analyses showed that the following factors significantly influenced the assessment: flaw size, shape, location and orientation, thermal hydraulic assumptions and material toughness. Applying national codes and procedures to the benchmark cases produced significantly different results in terms of allowable material toughness. This was mainly related to the safety factors used and the

  18. Experimental study of the effect of neutron radiation on pressurised water reactor vessel steel resilience - First part

    International Nuclear Information System (INIS)

    Verdeau, Jean-Jacques

    1969-12-01

    After having outlined the importance of the embrittlement of vessel steels by neutrons during the exploitation of pressurised water reactors, the author reports a set of tests which aimed at determining the effect of neutron irradiation on vessel steel resilience for operated, under construction or projected pressurized water reactors. He also tries to highlight the influence of irradiation temperature and of initial thermal treatments, and to look for a restoration thermal treatment of neutron-induced damages which could be applied to the considered vessels. Tests were performed on V Charpy resilience samples. Some samples have been irradiated by the Pile Department of the Grenoble CEN and then broken by the Laboratory of very high activity, whereas other samples have been irradiated in a prototype vessel and broken by a Cadarache department. The author presents characteristics of the studied steels (chemical compositions, thermal treatments), describes sample irradiation conditions, and the method of assessment of the transition temperature after irradiation, presents experimental results, discusses their interpretation, and presents future tests to be performed [fr

  19. Thermal Hydraulic Analysis of RPV Support Cooling System for HTGR

    International Nuclear Information System (INIS)

    Min Qi; Wu Xinxin; Li Xiaowei; Zhang Li; He Shuyan

    2014-01-01

    Passive safety is now of great interest for future generation reactors because of its reduction of human interaction and avoidance of failures of active components. reactor pressure vessel (RPV) support cooling system (SCS) for high temperature gas-cooled reactor (HTGR) is a passive safety system and is used to cool the concrete seats for the four RPV supports at its bottom. The SCS should have enough cooling capacity to ensure the temperature of the concrete seats for the supports not exceeding the limit temperature. The SCS system is composed of a natural circulation water loop and an air cooling tower. In the water loop, there is a heat exchanger embedded in the concrete seat, heat is transferred by thermal conduction and convection to the cooling water. Then the water is cooled by the air cooler mounted in the air cooling tower. The driving forces for water and air are offered by the density differences caused by the temperature differences. In this paper, the thermal hydraulic analysis for this system was presented. Methods for decoupling the natural circulation and heat transfer between the water loop and air flow were introduced. The operating parameters for different working conditions and environment temperatures were calculated. (author)

  20. Evaluation of fracture mechanics analyses used in RPV integrity assessment regarding brittle fracture

    International Nuclear Information System (INIS)

    Moinereau, D.; Faidy, C.; Valeta, M.P.; Bhandari, S.; Guichard, D.

    1997-01-01

    Electricite de France has conducted during these last years some experimental and numerical research programmes in order to evaluate fracture mechanics analyses used in nuclear reactor pressure vessels structural integrity assessment, regarding the risk of brittle fracture. These programmes included cleavage fracture tests on large scale cladded specimens containing subclad flaws with their interpretations by 2D and 3D numerical computations, and validation of finite element codes for pressurized thermal shocks analyses. Four cladded specimens made of ferritic steel A508 C13 with stainless steel cladding, and containing shallow subclad flaws, have been tested in four point bending at very low temperature in order to obtain cleavage failure. The specimen failure was obtained in each case in base metal by cleavage fracture. These tests have been interpreted by two-dimensional and three-dimensional finite element computations using different fracture mechanics approaches (elastic analysis with specific plasticity corrections, elastic-plastic analysis, local approach to cleavage fracture). The failure of specimens are conservatively predicted by different analyses. The comparison between the elastic analyses and elastic-plastic analyses shows the conservatism of specific plasticity corrections used in French RPV elastic analyses. Numerous finite element calculations have also been performed between EDF, CEA and Framatome in order to compare and validate several fracture mechanics post processors implemented in finite element programmes used in pressurized thermal shock analyses. This work includes two-dimensional numerical computations on specimens with different geometries and loadings. The comparisons show a rather good agreement on main results, allowing to validate the finite element codes and their post-processors. (author). 11 refs, 24 figs, 3 tabs

  1. Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

    International Nuclear Information System (INIS)

    Sokolov, Mikhail A; Lucon, Enrico

    2015-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 10 11 n/cm 2 /s (>1 MeV) to fluences from 0.5 to 3.4 10 19 n/cm 2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 10 13 n/cm 2 /s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 10 13 n/cm 2 /s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 10 19 n/cm 2 . The irradiation-induced shifts of the Master Curve reference temperatures, ΔT 0 , for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, ΔT 0 , 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT 0 , were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

  2. Evaluation of fracture mechanics analyses used in RPV integrity assessment regarding brittle fracture

    Energy Technology Data Exchange (ETDEWEB)

    Moinereau, D [Electricite de France, Dept. MTC, Moret-sur-Loing (France); Faidy, C [Electricite de France, SEPTEN, Villeurbanne (France); Valeta, M P [Commisariat a l` Energie Atomique, Dept. DMT, Gif-sur-Yvette (France); Bhandari, S; Guichard, D [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)

    1997-09-01

    Electricite de France has conducted during these last years some experimental and numerical research programmes in order to evaluate fracture mechanics analyses used in nuclear reactor pressure vessels structural integrity assessment, regarding the risk of brittle fracture. These programmes included cleavage fracture tests on large scale cladded specimens containing subclad flaws with their interpretations by 2D and 3D numerical computations, and validation of finite element codes for pressurized thermal shocks analyses. Four cladded specimens made of ferritic steel A508 C13 with stainless steel cladding, and containing shallow subclad flaws, have been tested in four point bending at very low temperature in order to obtain cleavage failure. The specimen failure was obtained in each case in base metal by cleavage fracture. These tests have been interpreted by two-dimensional and three-dimensional finite element computations using different fracture mechanics approaches (elastic analysis with specific plasticity corrections, elastic-plastic analysis, local approach to cleavage fracture). The failure of specimens are conservatively predicted by different analyses. The comparison between the elastic analyses and elastic-plastic analyses shows the conservatism of specific plasticity corrections used in French RPV elastic analyses. Numerous finite element calculations have also been performed between EDF, CEA and Framatome in order to compare and validate several fracture mechanics post processors implemented in finite element programmes used in pressurized thermal shock analyses. This work includes two-dimensional numerical computations on specimens with different geometries and loadings. The comparisons show a rather good agreement on main results, allowing to validate the finite element codes and their post-processors. (author). 11 refs, 24 figs, 3 tabs.

  3. Extension of the RPV irradiation surveillance program of NPP GKN II by T0 approach

    International Nuclear Information System (INIS)

    Barthelmes, J.; Keim, E.; Hein, H.; Koenig, G.

    2015-01-01

    The nuclear power plant (NPP) Neckarwestheim II (GKN II) started operation in 1989 and was designed for 40 years of operation. During the plant life time the reactor pressure vessel (RPV) integrity is a main aspect for nuclear safety since the RPV is exposed to neutron irradiation affecting the mechanical material properties, in particular toughness. In this context the ductile to brittle transition reference temperature of the RPV materials can be determined either indirectly according to the RT(NDT) concept by means of comparative examinations of irradiated and unirradiated notched-bar impact specimens or directly according to the Master Curve concept by means of examination of irradiated fracture mechanic specimens and determination of an alternative reference temperature RT(T0). With the implementation and evaluation of the first irradiation surveillance program consisting of three sets, one unirradiated reference set (set 1) and two irradiated sets (set 2 and 3), the RPV safety could be proven for the assessment fluence (AF) of 8*10 18 cm -2 (E > 1 MeV) using the RT(NDT) concept. Against the background of a possible long term operation and the state-of-the-art of science and technology in 1998 the NPP GKN II initiated a supplemental irradiation surveillance program with two irradiation sets (set 4 and 5) containing fracture mechanic specimens for complementary proof of safety according to the Master Curve concept. The results of the first irradiated set 4 are presented and assessed by means of the reference temperatures according to the Master Curve concept and compared to the results of the irradiation sets 1 to 3 of the conventional irradiation surveillance program. As an important outcome the existing RPV integrity assessment could be ensured by the Master Curve results. The applied approach adapts to the state-of-the-art of science and technology and is best practice to ensure the safe operation of RPV supplementary. (authors)

  4. Post mortem investigations of the NPP Greifswald WWER-440 reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Schuhknecht, Jan; Viehrig, Hans-Werner; Weiss, Frank-Peter; Rindelhardt, Udo [Forschungszentrum Dresden-Rossendorf e.V., Dresden (Germany). Inst. for Safety Research; Keller, Werner [Studsvik GmbH und Co. KG, Stutensee (Germany)

    2008-07-01

    The paper presents first results of the post mortem investigations performed at the reactor pressure vessels (RPV) of the Russian WWER-440 type reactors. Trepans were taken from the core weld SN0.1.4 and base metal of the unit 1 RPV. This RPV was annealed after 15 years of operation and operated for two more years. At first the trepan of the core welding seam was investigated by Master Curve (MC) testing. Specimens from 5 locations through the thickness of the welding seam were tested according to ASTM E1921-05. The reference temperature T{sub 0} was calculated with the measured fracture toughness values, K{sub Jc}, at brittle failure of the specimen. Generally the K{sub Jc} values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follow the course of the Master Curve. The K{sub Jc} values show a remarkable scatter. More values than expected lie below the 5% fractile. In addition the MC SINTAP procedure was applied to determine T{sub 0}{sup SINTAP} of the brittle fraction of the data set. There are remarkable differences between T{sub 0} and T{sub 0}{sup SINTAP} indicating macroscopic inhomogeneous weld metal. The highest T{sub 0} was about 50 C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T{sub 0} at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material may not represent the most conservative condition. The results presented in this paper show that the Master Curve approach as adopted in the test standard ASTM E1921-05 is applicable to the investigated WWER-440 multilayer weld metal. The results are of direct importance for an advanced WWER-440 RPV integrity assessment. On the other hand the data pool is broadened for a general introduction of the MC based RPV integrity assessment in the national codes. Furthermore general experiences in

  5. Post mortem investigations of the NPP Greifswald WWER-440 reactor pressure vessels

    International Nuclear Information System (INIS)

    Schuhknecht, Jan; Viehrig, Hans-Werner; Weiss, Frank-Peter; Rindelhardt, Udo

    2008-01-01

    The paper presents first results of the post mortem investigations performed at the reactor pressure vessels (RPV) of the Russian WWER-440 type reactors. Trepans were taken from the core weld SN0.1.4 and base metal of the unit 1 RPV. This RPV was annealed after 15 years of operation and operated for two more years. At first the trepan of the core welding seam was investigated by Master Curve (MC) testing. Specimens from 5 locations through the thickness of the welding seam were tested according to ASTM E1921-05. The reference temperature T 0 was calculated with the measured fracture toughness values, K Jc , at brittle failure of the specimen. Generally the K Jc values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follow the course of the Master Curve. The K Jc values show a remarkable scatter. More values than expected lie below the 5% fractile. In addition the MC SINTAP procedure was applied to determine T 0 SINTAP of the brittle fraction of the data set. There are remarkable differences between T 0 and T 0 SINTAP indicating macroscopic inhomogeneous weld metal. The highest T 0 was about 50 C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T 0 at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material may not represent the most conservative condition. The results presented in this paper show that the Master Curve approach as adopted in the test standard ASTM E1921-05 is applicable to the investigated WWER-440 multilayer weld metal. The results are of direct importance for an advanced WWER-440 RPV integrity assessment. On the other hand the data pool is broadened for a general introduction of the MC based RPV integrity assessment in the national codes. Furthermore general experiences in the cutting of irradiated RPV steels are collected

  6. Studies on the welding of heavy-section ASTM A542 Cl. 1 steel for large-sized pressure vessels

    International Nuclear Information System (INIS)

    Shimizu, Shigeki; Aota, Toshiichi; Kasahara, Masayuki

    1977-01-01

    ASTM A 542, Cl. 1 steel was developed and standardized recently, and is excellent in the high temperature strength and toughness as compared with conventionally used A 387, Grade 22 steel, accordingly the application to large pressure vessels is planned. This steel is a low alloy steel, and in case of large thickness, the possibility of cracking in the welded part is large. Also many times of annealing are required for the prevention of welding cracking, the relieving of residual stress, and the softening of hardened portion, but the possibility of cracking during stress-relieving annealing is large. In this study, Tekken type cracking test was carried out by coated electrode welding, and restricted cracking test was carried out by submerged arc welding of the A 542, Cl. 1 steel and A 387, Grade 22 steel, thus the welding cracking property was investigated, and the optimal welding conditions were selected. Also the test of cracking during the stress-relieving annealing of both steels was carried out, and the method of preventing the cracking was studied. The optimal conditions of stress-relieving annealing were selected, and the mechanism of the cracking was clarified. The mechanical properties of the joints welded and stress-relieved under the selected conditions were confirmed. (Kako, I.)

  7. Strain measurement and analysis for the RPV of Qinshan NPP (unit I) at primary system hydrostatic test

    International Nuclear Information System (INIS)

    Qu Jiadi; Wang Peizhu; Xie Shiqiu; Chen Renchang; Sheng Xianke; Dou Yikang; Zhao Weiliang

    1994-01-01

    Hydrostatic test for RPV (Reactor Pressure Vessel) is not only a means to inspect the vessels and the associated systems but also an important way to verify the results of mechanical analysis. The loading obtained by measurement is useful for the establishment of loading spectrum. Some discussions on the shop hydrostatic test planning for the RPV of Qinshan NPP (Nuclear Power Plant) performed in Japan are presented. Comparisons between the results of hydrostatic test provided by vendor and those of primary system hydrostatic test conducted at Qinshan Site are also given. Some data obtained at Qinshan Site such as actual loading and technical data of the stud-bolt, are listed. The results of measurement for the flange rotation, important for the sealing characteristics of RPV, are specifically discussed. The authors point out some of the mistakes in the results of the shop hydrostatic test

  8. The prediction of failure of welded steel plates for pressure vessels

    International Nuclear Information System (INIS)

    Gulvin, T.F.; Terry, P.; Webster, S.E.

    1980-01-01

    The avoidance of brittle fracture in pressure vessels and other welded steel fabrications is a matter of considerable importance. The application of fracture mechanics to this problem has led to the evolution of a philosophy which, among other things, permits estimation of tolerable crack sizes so that practical working limits can be set on inspection procedures and on material and welded joint toughness levels. The use of small-scale fracture mechanics tests, particularly crack opening displacement (COD) tests, to make the necessary material and/or joint property assessments is now commonplace. A number of possible analytical techniques have been considered. Initially, the COD data have been used with the Burdekin-Dawes design curve approach taking into account all the possible stress conditions and, it can be demonstrated, that consistently safe predictions of allowable flaw sizes can be made. Also considered, is the fracture analysis diagram approach of Harrison, Loosemore, and Milne in which a combination of fracture toughness data and mechanical property data is used to assess the probability of failure. Similarly the approach defined by Irvine and Quirk which makes use of mechanical property data without resorting to fracture mechanics and finally, a simple approach using uniaxial tensile property data only has also been examined. Comparisons have been made between these four approaches using, initially, the body of data referring to fracture tests on a single material with various defect sizes and aspect ratios, etc. and latterly, to the specific cases in the body of data on high strength steels. The results are discussed. (author)

  9. Critical experiments, measurements, and analyses to establish a crack arrest methodology for nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Hahn, G.T.

    1977-01-01

    Substantial progress was made in three important areas: crack propagation and arrest theory, two-dimensional dynamic crack propagation analyses, and a laboratory test method for the material property data base. The major findings were as follows: Measurements of run-arrest events lent support to the dynamic, energy conservation theory of crack arrest. A two-dimensional, dynamic, finite-difference analysis, including inertia forces and thermal gradients, was developed. The analysis was successfully applied to run-arrest events in DCB (double-cantilever-beam) and SEN (single-edge notched) test pieces. A simplified procedure for measuring K/sub D/ and K/sub Im/ values with ordinary and duplex DCB specimens was demonstrated. The procedure employs a dynamic analysis of the crack length at arrest and requires no special instrumentation. The new method was applied to ''duplex'' specimens to measure the large K/sub D/ values displayed by A533B steel above the nil-ductility temperature. K/sub D/ crack velocity curves and K/sub Im/ values of two heats of A533B steel and the corresponding values for the plane strain fracture toughness associated with static initiation (K/sub Ic/), dynamic initiation (K/sub Id/), and the static stress intensity at crack arrest (K/sub Ia/) were measured. Possible relations among these toughness indices are identified. During the past year the principal investigators of the participating groups reached agreement on a crack arrest theory appropriate for the pressure vessel problem. 7 figures

  10. 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

  11. Effects of thermal ageing on toughness properties of pressure vessel steel

    International Nuclear Information System (INIS)

    Todeschini, P.; Churier-Bossennec, H.; Massoud, J.P.; Frund, J.M.

    2015-01-01

    The reactor pressure vessel of pressurized water reactors operates at temperatures up to 325 C. degrees. The compositions and microstructures of its constitutive steel are optimized to obtain good initial toughness values and to minimize the effects of thermal ageing during service life. Intergranular segregation of embrittling elements like phosphorus is the main thermal ageing mechanism which might affect the long term toughness properties of low copper steels, despite the low diffusivity of phosphorus at the temperatures of interest. For long term operation, these effects are taken into account by prediction formulae which have been developed in the eighties and are included in the RCC-M and RSE-M codes. The presented study aims at validating these prediction formulae by exposures at moderately increased temperatures, up to 350 C. degrees, relatively to service conditions. The investigated materials are representative forgings and their welds, taking into account envelope phosphorus concentrations relatively to the French fleet. Predicted and measured embrittlement for base and weld metals are low and consistent together for the lowest phosphorus levels. The predicted effect of phosphorus content seems to be overestimated. The single coarse grain structure has been studied on one forging and shows a susceptibility to ageing similar to the fine grain one. The various heat affected zone microstructures studied with the plate having a phosphorus content of 0.017 % (fusion line, fine grains, inter-critical coarse grains) have given quite contrasted results. Inter-critical coarse grains notch positions show the lowest shifts. Code predictions are bounding the results of all considered heat affected zone microstructures with substantial margin. The increased susceptibility of heat affected zone compared to base metal seems globally overestimated

  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. The response of pressure vessel steel specimens on drop weight loading

    International Nuclear Information System (INIS)

    Winkler, S.; Kalthoff, J.F.; Gerscha, A.

    1979-01-01

    Load records obtained in instrumented impact tests in general are disturbed by inertia effects. The influence of mechanical damping provisions on these disturbing inertia effects is investigated. Precracked bend specimens are dynamically loaded in a drop weight testing system. The specimens of size 620 mm x 150 mm (25 mm or 50 mm thick) were machined from the pressure vessel steel 22 NiMoCr 37 which was heat treated to achieve a specially hardened condition. The tests were performed at two different low temperatures. The impact velocity was about 4 m/s. As it is usual in instrumented impact testing, the load at the tup of the impining striker is recorded as a function of time during the impact process. In addition the specimen is instrumented by a strain gage close to the crack tip in order to directly measure the stress intensification. Experiments were performed under pure and damped impact conditions. Damping was achieved by utilizing a soft aluminum plate between the striker and the specimen. (orig.)

  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. Appropriate welding conditions of temper bead weld repair for SQV2A pressure vessel steel

    International Nuclear Information System (INIS)

    Mizuno, R.; Matsuda, F.; Brziak, P.; Lomozik, M.

    2004-01-01

    Temper bead welding technique is one of the most important repair welding methods for large structures for which it is difficult to perform the specified post weld heat treatment. In this study, appropriate temper bead welding conditions to improve the characteristics of heat affected zone (HAZ) are studied using pressure vessel steel SQV2A corresponding to ASTM A533 Type B Class 1. Thermal/mechanical simulator is employed to give specimens welding thermal cycles from single to quadruple cycle. Charpy absorbed energy and hardness of simulated CGHAZ by first cycle were degraded as compared with base metal. Improvability of these degradations by subsequent cycles is discussed and appropriate temper bead thermal cycles are clarified. When the peak temperature lower than Ac1 and near Ac1 in the second thermal cycle is applied to CGAHZ by first thermal cycle, the characteristics of CGHAZ improve enough. When the other peak temperatures (that is, higher than Ac1) in the second thermal cycle are applied to the CGHAZ, third or more thermal cycle temper bead process should be applied to improve the properties. Appropriate weld condition ranges are selected based on the above results. The validity of the selected ranges is verified by the temper bead welding test. (orig.)

  16. The elevated temperature and thermal shock fracture toughnesses of nuclear pressure vessel steel

    International Nuclear Information System (INIS)

    Hirano, Kazumi; Kobayashi, Hideo; Nakazawa, Hajime; Nara, Atsushi.

    1979-01-01

    Thermal shock experiments were conducted on nuclear pressure vessel steel A533 Grade B Class 1. Elastic-plastic fracture toughness tests were carried out within the same high temperature range of the thermal shock experiment and the relation between stretched zone width, SZW and J-integral was clarified. An elastic-plastic thermal shock fracture toughness value. J sub(tsc) was evaluated from a critical value of stretched zone width, SZW sub(tsc) at the initiation of thermal shock fracture by using the relation between SZW and J. The J sub(tsc) value was compared with elastic-plastic fracture toughness values, J sub( ic), and the difference between the J sub(tsc) and J sub( ic) values was discussed. The results obtained are summarized as follows; (1) The relation between SZW and J before the initiation of stable crack growth in fracture toughness test at a high temperature can be expressed by the following equation regardless of test temperature, SZW = 95(J/E), where E is Young's modulus. (2) Elevated temperature fracture toughness values ranging from room temperature to 400 0 C are nearly constant regardless of test temperature. It is confirmed that upper shelf fracture toughness exists. (3) Thermal shock fracture toughness is smaller than elevated temperature fracture toughness within the same high temperature range of thermal shock experiment. (author)

  17. Studies of fragileness in steels of vessels of BWR reactors; Estudios de fragilizacion en aceros de vasija de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.F.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The structural materials with those that are manufactured the pressure vessels of the BWR reactors, suffer degradation in its mechanical properties mainly to the damage taken place by the fast neutrons (E > 1 MeV) coming from the reactor core. Its are experimentally studied those mechanisms of neutron damage in this material type, by means of the irradiation of steel vessel in experimental reactors to age them quickly. Alternatively it is simulated the neutron damage by means of irradiation of steel with heavy ions. In this work those are shown first results of the damage induced by irradiation from a similar steel to the vessel of a BWR reactor. The irradiation was carried out with fast neutrons (E > 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA MARK lll reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator, E = 4.8 MeV and range of the ionic flow of 0.1 to 53 iones/A{sup 2}. (Author)

  18. Comparison of SA508 Gr.3 and SA508 Gr.4N Low Alloy Steels for Reactor Pressure Vessel Steel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S

    2009-12-15

    The microstructural characteristics and mechanical properties of SA508 Gr.3 Mn-Mo-Ni low alloy steel and SA508 Gr.4N Ni-Mo-Cr low alloy steel were investigated. The differences in the stable phases between these two low alloy steels were evaluated by means of a thermodynamic calculation using ThermoCalc. They were then compared to microstructural features and correlated with mechanical properties. Mn-Mo-Ni low alloy steel shows the upper bainite structure which has the coarse cementite in the lath boundaries. However, Ni-Mo-Cr low alloy steel shows the mixture of lower bainite and tempered martensite structure that homogeneously precipitates the small carbides such as M{sub 23}C{sub 6} and M{sub 7}C{sub 3} due to an increase of hardenability and Cr addition. In the mechanical properties, Ni-Mo-Cr low alloy steel has higher strength and toughness than Mn-Mo-Ni low alloy steel. Ni and Cr additions increase the strength by solid solution hardening. Besides, microstructural changes from upper bainite to tempered martensite improve the strength of the low alloy steel by grain refining effect. And the changes in the precipitation behavior by Cr addition improve the ductile-brittle transition behavior along with a toughening effect of Ni addition.

  19. Integrity assessment of TAPS reactor pressure vessel at extended EOL using surveillance test results

    International Nuclear Information System (INIS)

    Chatterjee, S.; Shah, Priti Kotak

    2008-05-01

    Integrity assessment of pressure vessels of nuclear reactors (RPV) primarily concentrates on the prevention of brittle failure and conditions are defined under which brittle failure can be excluded. Accordingly, two approaches based on Transition Temperature Concept and Fracture Mechanics Concept were adopted using the impact test results of three credible surveillance data sets obtained from the surveillance specimens of Tarapur Atomic Power Station. RT NDT data towards end of life (EOL) were estimated from the impact test results in accordance with the procedures of USNRC Regulatory Guide 1.99, Rev. 2 and were used as primary input for assessment of the vessel integrity. SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station (TAPS). The reactors were commissioned during the year 1969. The chemical compositions of SA302B (modified) steel used in fabricating the vessel and the specified tensile property and the Charpy impact property requirements of the steel broadly meet ASME specified requirements. Therefore, the pressure temperature limit curves prescribed by General Electric (G.E.) were compared with those as obtained using procedures of ASME Section XII, Appendix G. The tensile and the Charpy impact properties at 60 EFPY of vessel operation as derived from the surveillance specimens even fulfilled the specified requirements for the virgin material of ASME. Integrity assessment carried out using the two approaches indicated the safety of the vessel for continued operation up to 60 EFPY. (author)

  20. Structural Integrity Assessment of VVER-1000 RPV under Accidental Cool down Transients

    International Nuclear Information System (INIS)

    Shrivastav, V.; Sen, R.N.; Yadav, R.S.

    2012-01-01

    Corrosion, Fatigue and Irradiation embrittlement are the major degradation mechanisms responsible for ageing of RPV (and its internals) of a Pressurized Water Reactor. While corrosion and fatigue can generate cracks, irradiation damage can lead to brittle fracture initiating from these cracks. Ageing in nuclear power plants needs to be managed so as to ensure that design functions remain available throughout the life of the plant. From safety perspective, this implies that ageing degradation of systems, structures and components important to safety remain within acceptable limits. Reactor Pressure Vessel has been identified as the highest priority key component in plant life management for Pressurized Water Reactors. Therefore special attention is required to ensure its structural integrity during its lifetime. In this paper, structural integrity assessment for typical VVER-1000 RPV is carried out under severe accidental cool down transients using the Finite Element Method. Three different accidental scenarios are postulated and safety of the vessel is conservatively assessed under these transients using the Linear Elastic Fracture Mechanics approach. Transient thermo mechanical stress analysis of the core belt region of the RPV is carried out in presence of postulated cracks and stress intensity factors are calculated and compared with the material fracture toughness to assess the structural integrity of the vessel. The paper also include some parametric analyses to justify the methodology. (author)

  1. Comparative study on two different seal surface structure for reactor pressure vessel sealing behavior

    International Nuclear Information System (INIS)

    Chen Jun; Xiong Guangming; Deng Xiaoyun

    2014-01-01

    The seal surface structure is very important to reactor pressure vessel (RPV) sealing behavior. In this paper, two 3-D RPV sealing analysis finite models have been established with different seal surface structures, in order to study the influence of two structures. The separation of RPV upper and lower flanges, bolt loads and etc. are obtained, which are used to evaluate the sealing behavior of the RPV. Meanwhile, the comparative analysis of safety margin of two seal surface structural had been done, which provides the theoretical basis for RPV seal structure design optimization. (authors)

  2. Study of Irradiation Effects on the Fracture Properties of A533-Series Ferritic Steels

    International Nuclear Information System (INIS)

    Lee, Yong Bok; Lee, Gyeong Geun; Kwon, Jun Hyun

    2011-01-01

    Since the Kori nuclear power plant unit 3 (Kori-3) was founded in 1986, the surveillance tests have been conducted five times. One of the primary objectives of the surveillance test is to determine the effects of irradiation on reactor pressure vessel (RPV) steel embrittlement. The RPV is made out of ferritic steels such as SA533 type B class 1, which were used for early nuclear power plants industry including Kori-2, 3, 4 and Yonggwang-1, 2 units in Korea. The Westinghouse supplied Kori-3 with the RPV steels ASTM A533 grade B class 1, which is equivalent to SA533 type B class 1. The irradiation effects on tensile properties in ASTM A533 grade B class 1 steel had been studied by Steichen and Williams. They experimentally determined the effect of strain rate and temperature on the tensile properties of unirradiated and irradiated A533 grade B steel 1. The effects of neutron irradiation on ferritic steels could be determined from tensile properties, as well as the fracture strength and toughness measurements. Hunter and Williams have reported that the strength and ductility for unirradiated material at a low strain rate increase with decreasing test temperature. Also, neutron irradiation increases strength and decreases ductility. Crosley and Ripling revealed that the yield strength of unirradiated material rapidly increases with the strain rate. Therefore, yield strength for unirradiated and irradiated materials should be determined by test parameters along with strain rate and temperature. In this study we compare ASTM A533 grad B class 1 steel obtained from several papers with SA533 type B class 1 steel taken from the surveillance data of Kori-3 unit, whose mechanical property of unirradiated and irradiated materials was correlated with the rate-temperature parameter

  3. Proof, interpretation and evaluation of radiation-induced microstructural changes in WWER reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Boehmert, J.; Gokhman, A.; Grosse, M.; Ulbricht, A.

    2003-06-01

    Neutron embrittlement is a special issue for the VVER-type reactors. One of the fundamentals for a reliable assessment of the current material state is knowledge of the causes and mechanisms of neutron embrittlement. The aim of the project is to understand and to quantify the microstructural appearances due to neutron radiation in VVER-type reactor pressure vessel steels. The material base is a broad variation of irradiation probes taken from the irradiation programme Rheinsberg, surveillance programmes of Russian, Ukrainian or Hungarian NPPs or irradiation experiments with mockup-alloys. The microstructure was investigated by different methods. The small angle neutron scattering (SANS) proved to be the most suitable method. A procedure was developed to determine mean diameter, size distribution and volume fraction of irradiation-induced microstructure from SANS experiments in a reliable and comparable manner. With this method microstructural parameters were systematically determined and the main factors of influence were identified. Apart from the neutron fluence the volume fraction of radiation defects mainly changes with the copper or nickel content whereas phosphorus is hardly relevant. Annealing remedies the radiation-induced microstructural appearances. The ratio between nuclear and magnetic neutron scattering provides information on the type of radiation defects. This leads to the conclusion that the material composition changes the radiation defects. The change occurs gradually rather than abruptly. The radiation defects detected by SANS correlate with the radiation hardening and embrittlement. Generally, the results suggest a bimodal mechanism due to radiation-enhanced and radiation-induced defect evolution. A kinetic model on base of the rate theory approach was established. (orig.)

  4. Experimental tests of irradiation-anneal-reirradiation effects on mechanical properties of RPV plate and weld materials

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1996-01-01

    The Charpy-V (C V ) notch ductility and tension test properties of three reactor pressure vessel (RPV) steel materials were determined for the 288 degree C (550 degree F) irradiated (I), 288 degree C (550 degree F) irradiated + 454 degree C (850 degree F)-168 h postirradiation annealed (IA), and 288 degree C (550 degree F) reirradiated (IAR) conditions. Total fluences of the I condition and the IAR condition were, respectively, 3.33 x 10 19 n/cm 2 and 4.18 x 10 19 n/cm 2 , E > 1 MeV. The irradiation portion of the IAR condition represents an incremental fluence increase of 1. 05 x 10 19 n/cm 2 , E > 1 MeV, over the I-condition fluence. The materials (specimens) were supplied by the Yankee Atomic Electric Company and represented high and low nickel content plates and a high nickel, high copper content weld deposit prototypical of the Yankee-Rowe reactor vessel. The promise of the IAR method for extending the fluence tolerance of radiation-sensitive steels and welds is clearly shown by the results. The annealing treatment produced full C V upper shelf recovery and full or nearly full recovery in the C V 41 J (30 ft-lb) transition temperature. The C V transition temperature increases produced by the reirradiation exposure were 22% to 43% of the increase produced by the first cycle irradiation exposure. A somewhat greater radiation embrittlement sensitivity and a somewhat greater reirradiation embrittlement sensitivity was exhibited by the low nickel content plate than the high nickel content plate. Its high phosphorus content is believed to be responsible. The IAR-condition properties of the surface vs. interior regions of the low nickel content plate are also compared

  5. Stress-relieving annealing of Cr-Mo steel for high temperature pressure vessels and the quality change in use

    International Nuclear Information System (INIS)

    Makioka, Minoru; Hirano, Hiromichi

    1976-01-01

    The securing of good mechanical properties is difficult in thick plates for large pressure vessels because cooling rate is insufficient and time is prolonged in heat treatment. Cr-Mo steel plates are usually used in the state of improved notch toughness though somewhat reduced strength by normalizing or accelerated cooling and tempering. If the time for heat treatment is prolonged, the embrittlement occurs. The effects of temperature, holding time, and cooling rate in stress-relieving treatment on the mechanical properties of 1-1/4Cr - 1/2Mo, 2-1/4Cr - 1Mo, 3Cr - 1Mo, and 5Cr - 1/2Mo steels were investigated. The tensile strength lowered almost linearly as the hollomon-Jaffe parameter of heat treatment condition increased in all the steels. The transition temperature shifted continuously to high temperature side in 1-1/4Cr - 1/2Mo steel, but the notch toughness was improved up to certain values and then the tendency turning to brittleness was shown in the other steels, as the H-J parameter increased. When the holding time became longer, the transition temperature shifted to higher temperature side, but the cooling rate showed no effect. The condition for stress relieving treatment must be selected so that the ferrite bands observed in welded metal do not arise. The embrittlement at the operation temperature of 400 - 450 0 C for a long time is evaluated by the comparison with that by stepped cooling method. (Kako, I.)

  6. Numerical simulation of a Charpy test and correlation of fracture toughness with fracture energy. Vessel steel and duplex stainless steel of the primary loop

    International Nuclear Information System (INIS)

    Breban, P; Eripret, C.

    1995-01-01

    The analysis methods used to evaluate the harmlessness of defects in the components of the primary coolant circuit of pressurized water reactor are based on the knowledge of the failure properties of concerned materials. The toughness is used to be measured through tests performed on normalized samples. But in some cases, especially for the vessel steel submitted to irradiation effects or for cast components in duplex stainless steel sensitive to thermal ageing, these measurements are not available on the material aged in operation. Therefore, fracture resistance has been evaluated through Charpy tests. Toughness is thus obtained on the basis of an empirical correlation. To improve these predictions, a modeling of the Charpy test in the framework of the local approach to fracture has been performed, for both materials. For the vessel steel, a complete evaluation of toughness has been achieved on the basis of a bidimensional viscoplastic modeling under large strain assumptions and a post-treatment with a Weibull model (cleavage fracture). The main hypothesis (partition between plain stress and plain strain areas in the bidimensional modeling) was corrected after a three dimensional calculations with the finite element program Code-Aster. The fracture analysis put into evidence that damage considerations like cavity nucleation and growth have to be introduced in the model in order to improve the description of physical phenomena. Two ways of progress have been suggested and are in course of being investigated, one in the framework of local approach to failure, the other with the help of micro-macro relationship. With regard to the duplex steel, the description of a Charpy (U) test allowed to clearly discriminate between crack initiation and propagation phases. A modeling through an equivalent homogenous material with a damage law based on a modified Gurson potential enables to describe quantitatively both phases of fracture. It clearly appears that a reliable

  7. Analysis of the necessity for inserting new surveillance capsule into the Kori Unit 1 RPV to monitor material fracture toughness

    International Nuclear Information System (INIS)

    Song, Taek Ho

    2007-01-01

    In association with monitoring of reactor pressure vessel (RPV) fracture toughness, surveillance capsule test specimens have been used to monitor the material property of nuclear reactor vessel. As far as Kori Unit 1 is concerned, 6 capsules were put into the vessel before commercial operation of the plant. Up to now, all the six capsules have been withdrawn to test and monitor the fracture toughness of RPV material. The last capsule has been withdrawn on June this year, and the Kori unit 1 has been shut downed since July 2007 and will be shut downed until December this year for about 6 months, preparing the life extension of the plant to operate the plant 10 more years. With the situation that all the surveillance capsules have been withdrawn, public ask the following question, 'To extend the life of Kori Unit 1 more than 10 years, is it necessary to insert new surveillance capsules into the Kori Unit 1 to monitor RPV material fracture toughness?' In connection with this issue, planning project have been carried out since spring this year. In this paper, it is described that inserting new surveillance capsule into the Kori Unit 1 RPV has some meaning in some public acceptance point of view and is not necessary in material engineering point of view

  8. Measurement of the yield and tensile strengths of neutron-irradiated and post-irradiation recovered vessel steels with notched specimens

    International Nuclear Information System (INIS)

    Valiente, A.

    1996-01-01

    Tensile circumferentially notched bars are examined as test specimens for measuring the yield and tensile strengths of nuclear pressure vessel steels under several conditions of irradiation and temperature that a vessel can experience during its service life, including recovery post-irradiation treatment. For all the vessel steels, notch geometries and conditions explored, it has been found that notched specimens fail by plastic collapse, and simple formulae have been derived that allow the yield and tensile strengths to be determined from the yielding and plastic collapse load of a notched specimen. Values measured in this way show good agreement with those measured by the standard tensile test method. (orig.)

  9. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    International Nuclear Information System (INIS)

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in x 1.2 m x 17.1 cm thick [4 ft x 4 ft x 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the open-quotes mirrorclose quotes insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in x 2.1 in [10 ft x 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28 degrees C/hr [12.5, 25, and 50 degrees F/hr] as measured on the heated face. A peak temperature of 454 degrees C [850 degrees F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing

  10. RPV SUSY searches at ATLAS and CMS

    CERN Document Server

    Pettersson, Nora Emilia; The ATLAS collaboration

    2015-01-01

    Experimental searches for Supersymmetry (SUSY) at the Large Hadronic Collider (LHC) often assume R-Parity Conservation (RPC) to avoid proton decay. A consequence RPC is that it implies a stable SUSY-particle that cannot decay. The search strategies are strongly based on the hypothesize of weakly interacting massive particles escaping without detection - yielding missing transverse energy (MET) to the collision events. It is vital to explore all possibilities considering that no observation of SUSY has been made and that strong exclusions already have been placed on RPC-SUSY scenarios. Introducing individually baryon- and lepton-number violating couplings in R-Parity Violating (RPV) models would avoid rapid proton decay. The strong mass and cross-section exclusion set for RPC-SUSY are weaken if RPV couplings are allowed in the SUSY Lagrangian - as these standard searches lose sensitivity due to less expected MET. This talk aims to summarise a few of the experimental searches for both prompt and long-lived RPV ...

  11. Advance of investigation of irradiation embrittlement mechanism of nuclear reactor pressure vessel steels. History and future of irradiation embrittlement researches

    International Nuclear Information System (INIS)

    Ishino, Shiori

    2007-01-01

    The nuclear reactor pressure vessel is the most important component of LWR plants required to be safe. This paper describes contents of the title consisting of four chapters. The first chapter states the general theory of irradiation effects, irradiation embrittlement and decreasing of toughness, and some kinds of pressure vessel steels. The second chapter explains history of irradiation embrittlement investigations and the advance of research methods for experiments and calculation. The third chapter contains information of inner structure of irradiated materials and development of prediction equations, recent information of embrittlement mechanism and mechanism guided prediction method, USA model and Central Research Institute of Electric Power Industry (CRIEPI) model. The fourth chapter states recent problems from viewpoints of experimental and analytical approaches. Comparison of standards of LWR pressure vessel steels between Japan and USA, relation between the density of number of cluster and the copper content, effect of flux on clustering of copper atoms, and CRIEPI's way of approaching the prediction method are illustrated. (S.Y.)

  12. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, A., E-mail: alan.peacock@ipp.mpg.de [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Girlinger, A. [MAN Diesel and Turbo SE D-94469 Deggendorf (Germany); Vorkoeper, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 17491 Greifswald (Germany); Boscary, J.; Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Hurd, F. [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany)

    2011-10-15

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m{sup 2}, are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m{sup 2} and a maximum heat load of 200 kW/m{sup 2} with a water velocity of approximately 2 m s{sup -1}. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m{sup 2} for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  13. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    International Nuclear Information System (INIS)

    Peacock, A.; Girlinger, A.; Vorkoeper, A.; Boscary, J.; Greuner, H.; Hurd, F.; Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G.

    2011-01-01

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m 2 , are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m 2 and a maximum heat load of 200 kW/m 2 with a water velocity of approximately 2 m s -1 . High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m 2 for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  14. The behavior of shallow flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Rolfe, S.T.

    1991-11-01

    Both analytical and experimental studies have shown that the effect of crack length, a, on the elastic-plastic toughness of structural steels is significant. The objective of this report is to recommend those research investigations that are necessary to understand the phenomenon of shallow behavior as it affects fracture toughness so that the results can be used properly in the structural margin assessment of reactor pressure vessels (RPVs) with flaws. Preliminary test results of A 533 B steel show an elevated crack-tip-opening displacement (CTOD) toughness similar to that observed for structural steels tested at the University of Kansas. Thus, the inherent resistance to fracture initiation of A 533 B steel with shallow flaws appears to be higher than that used in the current American Society of Mechanical Engineers (ASME) design curves based on testing fracture mechanics specimens with deep flaws. If this higher toughness of laboratory specimens with shallow flaws can be transferred to a higher resistance to failure in RPV design or analysis, then the actual margin of safety in nuclear vessels with shallow flaws would be greater than is currently assumed on the basis of deep-flaw test results. This elevation in toughness and greater resistance to fracture would be a very desirable situation, particularly for the pressurized-thermal shock (PTS) analysis in which shallow flaws are assumed to exist. Before any advantage can be taken of this possible increase in initiation toughness, numerous factors must be analyzed to ensure the transferability of the data. This report reviews those factors and makes recommendations of studies that are needed to assess the transferability of shallow-flaw toughness test results to the structural margin assessment of RPV with shallow flaws. 14 refs., 8 figs

  15. Corrosion properties of sealing surface material for RPV under abnormal working conditions

    International Nuclear Information System (INIS)

    Liu Jinhua; Wen Yan; Zhang Xuemei; Hou Songmin; Gong Bin; He Yanchun

    2012-01-01

    Based on the corrosion issue of sealing surface material for RPV in some nuclear projects, the corrosion properties of sealing surface material for RPV under abnormal working conditions were investigated. The corrosion behavior of 308L stainless steel were studied by using autoclave in different contents of Cl - solutions, and these samples were observed and analyzed by means of the metalloscope and Scanning electron microscope (SEM). Results show that no pitting, crevice and stress corrosion occurred, when the content of Cl - was lower than 1 mg/L at the temperatures of 270℃ and the pressure of 5.5 MPa. However, with the increase of the content of Cl - , the susceptibility to pitting, crevice and stress corrosion of 308L was enhanced remarkably. (authors)

  16. Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

    Energy Technology Data Exchange (ETDEWEB)

    Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J. [Centre de l``Etude de l``Energie Nucleaire, Mol (Belgium); Biemiller, E.C. [Yankee Atomic Electric Company, Bolton (United States); Rossinski, S.T.; Carter, R.G. [Electric Power Research Institute, Charlotte (United States)

    1996-07-01

    The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

  17. Embrittlement and annealing of reactor pressure vessel steels: comparison of BR3 surveillance and vessel plates to the surrogate plates representative of the Yankee Rowe vessel

    International Nuclear Information System (INIS)

    Fabry, A.; Chaouadi, T.; Puzzolante, J.L.; Van de Velde, J.; Biemiller, E.C.; Rossinski, S.T.; Carter, R.G.

    1996-07-01

    The sister pressure vessels at the BR3 and Rowe Yankee PWR plants were operated at a lower-than-usual temperature (260 degrees Celsius) and their plates were austenitized at higher-than-usual temperature (970 degrees Celsius). A heat tratemement leading to a coarser microstructure than typical for the fine grain plates that are considered in development of USNRC Regulatory guide 1.99. This material displayed outlier behaviour charackterized by a 41J CVN-shift significantly larger than predicted by Regulatory Guide 1.99. Because lower radiation temperature and nickell alloying are generally considered detrimental to irradiation sensitivity, there was a major concern that the nickel-modified lower Rowe plate and the nickel-modified BR3 plate may become too embrittled to satisfy the toughness requirements enbodied in the PTS screening criterion. This paper compares three complementary studies undertaken to clarify these uncertainties: 1) the accelerated irradiation and test program launched in 1990 by Yankee Atomic Electric Company using typical vessel plate materials containing 0.24% copper at two nickel levels: YA1, 0.63% (A533-B) and YA9, 0.19% (A302-B). These were heat-treated to produce the coarse and fine grain microstructures representative of the Yankee/BR3 and the Regulatory Guide plates, respectively, 2) the BR3 surveillance and vessel testing program: this vessel was wet-annealed in 1984, relicensed for operation till the plant shutdown in 1987, ANCL was trepanned in early 1995, 3) the accelerated irradiations in the Belgian BR2 test reactor of the Yankee coarse grain plates YA1 and YA9 together with BR3 vessel specimens extracted at nozzle elevation, a location with negligible radiation exposure. It is shown that the PTS screening criterion was never attained by the BR3 and Rowe plates, and that the BR3 vessel anneal was neither necessary nor sufficient. Finally, the sensitivity of embrittlement, annealing and post-annealing reembrittlement to irradiation

  18. Heavy-section steel irradiation program. Volume 4, No. 2. Semiannual progress report, April 1993--September 1993

    International Nuclear Information System (INIS)

    Corwin, W.R.

    1995-03-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents which have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. In particular, it is vital to fully understand the degree of irradiation-induced degradation of the RPV's fracture resistance which occurs during service, since without that radiation damage, it is virtually impossible to postulate a realistic scenario that would result in RPV failure. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established to provide a quantitative assessment of the effects of neutron irradiation on the material behavior and, in particular, the fracture toughness properties of typical pressure-vessel steels. Effects of specimen size; material chemistry; product form and microstructure; irradiation fluence, flux, temperature, and spectrum; and postirradiation annealing are being examined on a wide range of fracture properties. The HSSI Program is arranged into 14 tasks: (1) program management, (2) fracture toughness (K lc ) curve shift in high-copper welds, (3) crack-arrest toughness (K la ) curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K lc and K la curve shifts in low upper-shelf (LUS) welds, (6) annealing effects in LUS welds, (7) irradiation effects in a commercial LUS weld, (8) microstructural analysis of irradiation effects, (9) in-service aged material evaluations, (10) correlation monitor materials, (11) special technical assistance, (12) Japan Power Development Reactor steel examination, (13) technical assistance for Joint Coordinating Committee on Civilian Nuclear Reactor Safety (JCCCNRS) Working Groups 3 and 12, and (14) additional requirements for materials

  19. 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)

  20. Criterion of cleavage crack propagation and arrest in a nuclear PWR vessel steel

    International Nuclear Information System (INIS)

    Bousquet, Amaury

    2013-01-01

    The purpose of this PhD thesis is to understand physical mechanisms of cleavage crack propagation and arrest in the 16MND5 PWR vessel steel and to propose a robust predicting model based on a brittle fracture experimental campaign of finely instrumented laboratory specimens associated with numerical computations. First, experiments were carried out on thin CT25 specimens at five temperatures (-150 C, -125 C, -100 C, -7 C, -50 C). Two kinds of crack path, straight or branching path, have been observed. To characterize crack propagation and to measure crack speed, a high-speed framing camera system was used, combined with the development of an experimental protocol which allowed to observe CT surface without icing inside the thermal chamber and on the specimen. The framing camera (520 000 fps) has allowed to have a very accurate estimation of crack speed on the complete ligament of CT (∼ 25 mm). Besides, to analyse experiments and to study the impact of viscosity on the mechanical response around the crack tip, the elastic-viscoplastic behavior of the ferritic steel has been studied up to a strain rate of 104 s -1 for the tested temperatures.The extended Finite Element Method (X-FEM) was used in CAST3M FE software to model crack propagation. Numerical computations combine a local non linear dynamic approach with a RKR type fracture stress criterion to a characteristic distance. The work carried out has confirmed the form of the criterion proposed by Prabel at -125 C, and has identified the dependencies of the criterion on temperature and strain rate. From numerical analyzes in 2D and 3D, a multi-temperature fracture stress criterion, increasing function of the strain rate, was proposed. Predictive modeling were used to confirm the identified criterion on two specimen geometries (CT and compressive ring) in mode I at different temperatures. SEM observations and 3D analyzes made with optical microscope showed that the fracture mechanism was the cleavage associated

  1. Inspection and repair of reactor pressure vessel (RPV) internals

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.; Nicolai, M.

    1996-01-01

    The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

  2. Master curve approach to monitor fracture toughness of reactor pressure vessels in nuclear power plants

    International Nuclear Information System (INIS)

    2009-10-01

    A series of coordinated research projects (CRPs) have been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on reactor pressure vessel (RPV) steels. The purpose of the CRPs was to develop correlative comparisons to test the uniformity of results through coordinated international research studies and data sharing. The overall scope of the eighth CRP (CRP-8), Master Curve Approach to Monitor Fracture Toughness of Reactor Pressure Vessels in Nuclear Power Plants, has evolved from previous CRPs which have focused on fracture toughness related issues. The ultimate use of embrittlement understanding is application to assure structural integrity of the RPV under current and future operation and accident conditions. The Master Curve approach for assessing the fracture toughness of a sampled irradiated material has been gaining acceptance throughout the world. This direct measurement of fracture toughness approach is technically superior to the correlative and indirect methods used in the past to assess irradiated RPV integrity. Several elements have been identified as focal points for Master Curve use: (i) limits of applicability for the Master Curve at the upper range of the transition region for loading quasi-static to dynamic/impact loading rates; (ii) effects of non-homogeneous material or changes due to environment conditions on the Master Curve, and how heterogeneity can be integrated into a more inclusive Master Curve methodology; (iii) importance of fracture mode differences and changes affect the Master Curve shape. The collected data in this report represent mostly results from non-irradiated testing, although some results from test reactor irradiations and plant surveillance programmes have been included as available. The results presented here should allow utility engineers and scientists to directly measure fracture toughness using small surveillance size specimens and apply the results using the Master Curve approach

  3. Inelastic Cyclic Deformation Behaviors of Type 316H Stainless Steel for Reactor Pressure Vessel of Sodium-Cooled Fast Reactor at Elevated Temperatures

    International Nuclear Information System (INIS)

    Yoon, Ji-Hyun; Hong, Seokmin; Koo, Gyeong-Hoi; Lee, Bong-Sang; Kim, Young-Chun

    2015-01-01

    Type 316H stainless steel is a primary candidate material for a reactor pressure vessel of a sodium-cooled fast (SFR) reactor which is under development in Korea. The reactor pressure vessel for a SFR is subjected to inelastic deformation induced by cyclic thermal stress. Fully reversed cyclic testing and ratcheting testing at elevated temperatures were performed to characterize the inelastic cyclic deformation behaviors of Type 316H stainless steel at the SFR operating temperature. It was found that cyclic hardening of Type 316H stainless steel was enhanced, and the accumulation of ratcheting deformation of Type 316H stainless steel was retarded at around the SFR operating temperature. The results of the tensile testing and the microstructural investigation for dislocated structures after the inelastic deformation testing showed that dynamic strain aging affected the inelastic cyclic deformation behavior of Type 316 stainless steel at around the SFR operating temperature.

  4. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    International Nuclear Information System (INIS)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi

    2017-01-01

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t_8_/_5 (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

  5. Study on the welding continuous cooling transformation and weldability of SA508Gr4 steel for nuclear pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bai, Qingwei; Ma, Yonglin; Xing, Shuqing; Chen, Zhongyi [Inner Mongolia Univ. of Science and Technology, Baotou (China). School of Material and Metallurgy; Kang, Xiaolan [Baotou Vocational and Technical College (China)

    2017-02-15

    SA508Gr4 is a newly developed high-strength steel for nuclear reactor pressure vessels. Its welding characteristics remain largely unexplored. In this work, the simulated heat affected zone continuous cooling transformation (SH-CCT) diagram of SA508Gr4 steel was constructed and the high-temperature cooling phase compositions and the properties of the heat affected zone (HAZ) were characterized using dilatometry and microscopic tests. The results show that the phase transformation in the HAZ was divided into bainite and martensite transformation stages. When 4.6 ≤ t{sub 8/5} (the HAZ cooling time from 800 C to 500 C) ≤ 15 s, lath-shaped martensite was fully developed, resulting in extensive hardening and cold cracking in the HAZ, while the cooling time required to form the bainite completely exceeds 1 200 s. Thus, to improve weld quality, preheating to 196 C or higher is recommended.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The IAEA Specialists` Meeting on Cracking in LWR RPV Head Penetrations was held at the ASTM Headquarters, Philadelphia, Pennsylvania, on May 2-4, 1995. It was attended by 39 participants from 12 countries. The meeting was held in the framework of the IAEA International Working Group on Life Management of Nuclear Power Plants (IWG-LMNPP) and was organized and sponsored by the Oak Ridge National Laboratory and the U.S. Nuclear Regulatory Commission. The purpose of the meeting was to review experience in the field for ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. Presentations were aimed at achieving a better understanding of the behaviour of reactor component materials, providing guidance and recommendations to assure reliability and adequate performance, and proposing directions for further investigations. Refs, figs and tabs.

  7. Site ultrasonic measurement on RPV stud-bolt loading under hot transient of Qinshan NPP

    International Nuclear Information System (INIS)

    Qu Jiadi; Dou Yikang; Zhu Shiming

    1994-08-01

    It expounds that the key of solving thermal transient sealing problem is to obtain the thermal increment of stud-bolt loading. This loading, as a primary stress loading, is directly related to the bolt fatigue life and transient loading spectrum for vessel analysis. The fundamental works and main results of ultrasonic measurement on RPV stud-bolt loading on Qinshan site are also presented. The measuring capability has exceeded 1 m in length and temperature of 280 degree C, therefore, it is possible to be used in the field of NPP. The paper is the continuation of research work for sealing analysis and tests on the RPV (see SMiRT-9, 10)

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

    International Nuclear Information System (INIS)

    1995-01-01

    The IAEA Specialists' Meeting on Cracking in LWR RPV Head Penetrations was held at the ASTM Headquarters, Philadelphia, Pennsylvania, on May 2-4, 1995. It was attended by 39 participants from 12 countries. The meeting was held in the framework of the IAEA International Working Group on Life Management of Nuclear Power Plants (IWG-LMNPP) and was organized and sponsored by the Oak Ridge National Laboratory and the U.S. Nuclear Regulatory Commission. The purpose of the meeting was to review experience in the field for ensuring adequate performance of reactor pressure vessel (RPV) heads and penetrations. Presentations were aimed at achieving a better understanding of the behaviour of reactor component materials, providing guidance and recommendations to assure reliability and adequate performance, and proposing directions for further investigations. Refs, figs and tabs

  9. In- and ex-vessel flooding as part of the severe accident strategy in the KERENA reactor

    International Nuclear Information System (INIS)

    Levi, P.; Fischer, M.

    2011-01-01

    Currently, AREVA NP is finalizing the basic design of the KERENA reactor, an advanced boiling water reactor with a net electric output of about 1250 MWe. The safety concept in the KERENA reactor is founded on reliable active and passive systems for water supply and heat removal. The passive systems are based on simple physics and do not require operator action. Therefore, a severe accident (SA) with core damage, caused by the subsequent and multiple failures of the safety systems, has an extremely low probability. Despite this, the KERENA design is intended to involve measures that can limit and stop the progression of the severe accident which further reduces the frequency and extent of radioactive releases into the environment. These additional measures include in-vessel and ex-vessel flooding. Flooding is intended to remove the heat from the core or from the reactor pressure vessel (RPV) and transfer it into the containment. There the heat is removed by the active RHR (residual heat removal) system or by the passive CCCs (containment cooling condensers). Both flooding measures are passive and actuated independent of each other by different signals. The study shows that the in-vessel flooding is capable of arresting the core melt progression before a large molten pool can develop. In the unlikely event that the passive in-vessel flooding cannot be actuated or fails, the core will melt and relocate into the lower head of the RPV. In this case, as a further line of defense, decay heat removal can be achieved through the RPV wall into the water in the cavity. In order to assess whether the ex-vessel cooling can ensure RPV wall integrity a dedicated thermodynamics code has been developed which considers heat transfer from the molten corium pool into the RPV wall and the resulting wall ablation. As an input for the code the stratification behavior of the oxidic and metallic phase of the molten pool is examined. In the case of a light metallic phase on top, high heat

  10. 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)

  11. Experimental investigations on vessel-hole ablation during severe accidents

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.; Green, J.A.; Paladino, D.

    1997-12-01

    This report presents experimental results, and subsequent analyses, of scaled reactor pressure vessel (RPV) failure site ablation tests conducted at the Royal Institute of Technology, Division of Nuclear Power Safety (RIT/NPS). The goal of the test program is to reduce the uncertainty level associated with the phase-change-ablation process, and, thus, improve the characterization of the melt discharge loading on the containment. In a series of moderate temperature experiments, the corium melt is simulated by the binary oxide CaO-B 2 O 3 or the binary eutectic and non-eutectic salts NaNO 3 -KNO 3 , while the RPV head steel is represented by a Pb, Sn or metal alloys plate. A complementary set of experiments was conducted at lower temperatures, using water as melt and salted ice as plate material. These experiments scale well to the postulated prototypical conditions. The multidimensional code HAMISA, developed at RIT/NPS, is employed to analyze the experiments with good pre- and post-test predictions. The effects of melt viscosity and crust surface roughness, along with failure site entrance and exit frictional losses on the ablation characteristics are investigated. Theoretical concept was proposed to describe physical mechanisms which govern the vessel-hole ablation process during core melt discharge from RPV. Experimental data obtained from hole ablation tests and separate-effect tests performed at RIT/NPS were used to validate component physical models of the HAMISA code. It is believed that the hole ablation phenomenology is quite well understood. Detailed description of experiments and experimental data, as well as results of analyses are provided in the appendixes

  12. Installation method for the steel container and vessel of the nuclear heating reactor

    International Nuclear Information System (INIS)

    Chen Liying; Guo Jilin; Liu Wei

    2000-01-01

    The Nuclear Heating Reactor (NHR) has the advantages of inherent safety and better economics, integrated arrangement, full power natural circulation and dual vessel structure. However, the large thin container presents a new and difficult problem. The characteristics of the dual vessel installation method are analyzed with system engineering theory. Since there is no foreign or domestic experience, a new method was developed for the dual vessel installation for the 5 MW NHR. The result shows that the installation method is safe and reliable. The research on the dual vessel installation method has important significance for the design, manufacture and installation of the NHR dual vessel, as well as the industrialization and standardization of the NHR

  13. Heavy-Section Steel Irradiation Program. Volume 2, No. 1: Semiannual progress report, October 1990--March 1991

    International Nuclear Information System (INIS)

    Corwin, W.R.

    1994-07-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure-vessel steels as they relate to light-water reactor pressure-vessel integrity. The HSSI Program is arranged into nine tasks: (1) program management, (2) K ic curve shift in high-copper welds, (3) K ia curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K ic and K ia curve shifts in low upper-shelf (LUS) weld, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation, (8) in-service aged material evaluations, and (9) correlation monitor materials. During this period, additional analyses on the effects of precleavage stable ductile tearing on the toughness of high-copper welds 72W and 73W demonstrated that the size effects observed in the transition region are not due to substantial differences in ductile tearing behavior. Possible modifications to irradiated duplex crack-arrest specimens were examined to increase the likelihood of their successful testing. Characterization of a second batch of 72W and 73W welds was begun and results of the Charpy V-notch testing is provided. A review of literature on the annealing response of reactor pressure vessel steels was initiated

  14. 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

  15. Use of advanced inspection technology during the ISI of a US-RPV

    Energy Technology Data Exchange (ETDEWEB)

    Buxbaum, S R; Pond, R B [Baltimore Gas and Electric Co., MD (United States); Stone, R M

    1988-12-31

    The Reactor Pressure Vessel (RPV) maintains a unique place among nuclear steam supply system components because its failure is unacceptable. The assumption of incredibility of vessel failure is a US Nuclear Regulatory Commission (USNRC) requirement of plant design and operation. Therefore, accurate detection and characterization of vessel flaws are essential. In order to meet the needs for improved pressure vessel inspection, EPRI assisted in the development of the Ultrasonic Data Recording and Processing System (UDRPS). The EPRI NDE Center has supported the transfer to industry through demonstration and documentation of the original system capability and by assisting utilities in their initial applications. Baltimore Gas and Electric (BG and E) purchased a second generation UDRPS and has used the system during the 10 year ISI at the Calvert Cliffs Nuclear Plant, Units 1 and 2. This presentation deals with the BG and E applications and the EPRI NDE Center support provided before and during the Calvert Cliffs ISI applications. (author).

  16. Corrosion of steel tendons in concrete pressure vessels: review of recent literature and experimental investigations

    International Nuclear Information System (INIS)

    Griess, J.C.

    1978-01-01

    The fundamentals of localized corrosion are briefly discussed, and the literature concerning corrosion of carbon steel in aqueous environments, in particular the stress-corrosion cracking of carbon steels, is reviewed. The behavior of high strength steels in specific environments, including concrete and organic substances, is also summarized. The available information indicates that the corrosion of steels in correctly formulated concrete is minimal. Even appreciable concentrations of chloride, sulfate, sulfide, and nitrate salts can be tolerated in the concrete or grout without detrimental effects. Adherence to established standards in the preparation and application of grouts in tendon-bearing conduits should guarantee very long tendon lifetimes. Little is reported about the behavior of tendons in proprietary organic greases or waxes, but very good corrosion resistance is expected if the organic material remains intact. Stress-corrosion cracking tests performed with AISI 1080 steel tendon wires, using the constant-strain-rate method, produced results expected from data in the literature. Cracking was observed only in neutral or acid solutions containing hydrogen sulfide, in ammonium nitrate solutions, and possibly in a dilute solution of sodium bisulfite. General corrosion tests in water and in dilute solutions of sodium nitrate, chloride, or sulfate showed that oxygen was an important factor; corrosion was substantially greater when oxygen had free access to the solution than when access to oxygen was restricted. In the tests with oxygen the heaviest attack on the steel tendons was at the waterline of the solution

  17. Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

    International Nuclear Information System (INIS)

    Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

    1983-01-01

    This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along with applications to pressure vessel experiments

  18. Monitoring the aging of pressure vessel steels by TEP measurements: Advantages and current limitations of the method

    International Nuclear Information System (INIS)

    Kleber, X.; Saillet, S.

    2011-01-01

    The TEP (Thermoelectric Power or Seebeck coefficient) characterizes the ability of a material to generate an electrical potential difference when the material is subjected to a heat flux. It can be defined from the Seebeck effect, which manifests itself in a circuit formed by two different metals subjected to a temperature gradient. The origin of the thermoelectric power is, as the resistivity, due to electronic phenomena occurring at the atomic scale in relation to the crystallographic structure of the material. TEP measurements are used to characterize small microstructural changes at the scale of crystal defects. The high sensitivity of TEP makes it an excellent probe able of detecting small changes in the microstructural state, including precipitation, dissolution of alloying elements, hardening and recovery after deformation. It has been shown recently that the TEP of pressure vessels steels was sensitive to irradiation, making this measurement technique a potential candidate for monitoring the aging of the pressure vessel steel. However, the first measurements on Charpy specimens of the EDF monitoring program (Pressure Vessel Surveillance Program) showed a strong negative effect of specimen geometry on the accuracy that can be achieved. In this paper we show what the origins of these inaccuracies are. From numerical simulation and finite element model, we describe the roles of the thermal contact resistance as well as the influence of the geometry of the blocks device. A model is proposed to overcome these negative effects. We also show the effect of the presence of heterogeneities in the material on the TEP measurement, and the importance of their localization. Finally, solutions are proposed to improve the device for measuring TEP on PVSP Charpy specimens. (authors)

  19. Influence of specimen size/type on the fracture toughness of five irradiated RPV materials

    Energy Technology Data Exchange (ETDEWEB)

    Sokolov, Mikhail A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lucon, Enrico [National Inst. of Standards and Technology (NIST), Boulder, CO (United States)

    2015-01-01

    The Heavy-Section Steel Irradiation (HSSI) Program had previously irradiated five reactor pressure vessel (RPV) steels/welds at fast neutron fluxes of about 4 to 8 x 1011 n/cm2/s (>1 MeV) to fluences from 0.5 to 3.4 1019 n/cm2 and at 288 °C. The unirradiated fracture toughness tests were performed by Oak Ridge National Laboratory with 12.7-mm and 25.4-mm thick (0.5T and 1T) compact specimens, while the HSSI Program provided tensile and 5 x 10-mm three-point bend specimens to SCK-CEN for irradiation in the in-pile section of the Belgian Reactor BR2 at fluxes > 1013 n/cm2/s and subsequent testing by SCK-CEN. The BR2 irradiations were conducted at about 2 and 4 x 1013 n/cm2/s with irradiation temperature between 295 °C and 300 °C (water temperature), and to fluences between 6 and 10 x 1019n/cm2. The irradiation-induced shifts of the Master Curve reference temperatures, ΔT0, for most of the materials deviated from the embrittlement correlations much more than expected, motivating the testing of 5 x 10-mm three-point bend specimens of all five materials in the unirradiated condition to eliminate specimen size and geometry as a variable. Tests of the unirradiated small bend specimens resulted in Master Curve reference temperatures, T0, 25 °C to 53 °C lower than those from the larger compact specimens, meaning that the irradiation-induced reference temperature shifts, ΔT0, were larger than the initial measurements, resulting in much improved agreement between the measured and predicted fracture toughness shifts.

  20. Technology development and production of elongated shell for reactor vessel active zone of WWER-TOI project from steel 15Cr2NiMoVN class 1

    International Nuclear Information System (INIS)

    Shklyaev, S.Eh.; Titova, T.I.; Ratushev, D.V.; Shul'gan, N.A.; Eroshkin, S.B.; Durynin, V.A.; Efimov, S.V.; Dub, V.S.; Kulikov, A.P.; Romashkin, A.N.

    2015-01-01

    Production process for the elongated shell blank of the active zone of the reactor pressure vessel made from steel 15Cr2NiMoVN Class 1 with finished sizes Dext=4.655 mm, Dint=4.240 mm, H=4.910 mm (height for heat treatment – 5.750 mm) is presented. For the first time in Russia in production site of OMZ-Special steel LLC a unique elongated shell blank of the reactor vessel active zone was made from ingot 420.0 t for WWER-TOI project fully meeting the specified requirements in terms of metallurgical quality and set of service properties [ru

  1. Formation of microcracks during stress-relief annealing of a weldment in pressure vessel steel of type A508 C1 2

    International Nuclear Information System (INIS)

    Liljestrand, L.-G.; Oestberg, G.; Lindhagen, P.

    1978-01-01

    Crack formation in the heat-affected zones of heavy section weldments of type A 508 C1 2 pressure vessel steel during stress-relief annealing has been studied on an actual weldment and on simulated structures. Mechanical testing of the latter showed that stress relaxation of the order of magnitude occuring during stress-relief annealing can produce cracks of the same kind as occasionally found in weldments of pressure vessel steel. The primary cause is believed to be grain boundary sliding, possibly but not necessarily enhanced by impurities. (Auth.)

  2. Thermo-mechanical behaviour during encapsulation of glass in a steel vessel

    International Nuclear Information System (INIS)

    Nakhodchi, S.; Smith, D.J.; Thomas, B.G.

    2016-01-01

    Quantitative numerical simulations and qualitative evaluations are conducted to elucidate thermo-mechanical behaviour during pouring and solidification of molten glass into a stainless-steel cylindrical container. Residual stress and structural integrity in this casting/vitrification process is important because it can be used for long-term storage of high-level nuclear wastes. The predicted temperature and stress distributions in the glass and container agree well with previous measurements of the temperature histories and residual stresses. Three different thermal-stress models are developed using the finite-element method and compared. Two simple slice models were developed based on the generalized plane strain assumption as well as a detailed two-dimensional axi-symmetric model that adds elements according to the stages of pouring glass into the stainless steel container. The results reveal that mechanical interaction between the glass and the wall of the stainless steel container generates residual tensile stresses that approach the yield strength of the steel. Together, these results reveal important insights into the mechanism of stress generation in the process, the structural integrity of the product, and accuracy of the modelling-tool predictions. - Highlights: • Source of residual stresses in glass and stainless steel containers (canisters) is discussed. • Final residual stresses in both glass and container is quantified. • Simple models presented for simulation of complicated casting process. • Comparison between detailed and simple FE modeling.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

  4. Results of performance testing the Russian RPV temperature measurement probe used for annealing

    International Nuclear Information System (INIS)

    Nakos, J.T.; Selsky, S.

    1998-03-01

    This paper provides information on three (3) topics related to temperature measurements in an annealing procedure: (1) results of a series of experiments performed by CNIITMASH of the Russian consortium MOHT on their reactor pressure vessel (RPV) temperature measurement probe, (2) a discussion regarding uncertainties and errors in RPV temperature measurements, and (3) predictions from a thermal model of a spherical RPV temperature measurement probe. MOHT teamed with MPR Associates and was to perform the Annealing Demonstration Project (ADP) on behalf of the US Department of Energy, ESEERCo, EPRI, CRIEPI, Framatome, and Consumers Power Co. at the Midland plant. Experimental results show that the CNIITMASH probe errors are a maximum of about 27 C (49 F) during a 15 C/hr (27 F/hr) heat-up but only about 3 C (5.4 F) (0.6%) during the hold portion at 470 C (878 F). These errors are much smaller than those obtained from a similar series of experiments performed by Sandia National Laboratories (Sandia). The discussion about uncertainties and errors shows that results presented as a temperature difference provides a measure of the probe error. Qualitative agreement is shown between the model predictions, the experimental results of the CNIITMASH probe and the experimental results of a series of similar experiments performed by Sandia

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

    International Nuclear Information System (INIS)

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

    1996-07-01

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

  6. Ductile fracture toughness of heavy section pressure vessel steel plate. A specimen-size study of ASTM A 533 steels

    International Nuclear Information System (INIS)

    Williams, J.A.

    1979-09-01

    The ductile fracture toughness, J/sub Ic/, of ASTM A 533, Grade B, Class 1 and ASTM A 533, heat treated to simulate irradiation, was determined for 10- to 100-mm thick compact specimens. The toughness at maximum specimen load was also measured to determine the conservatism of J/sub Ic/. The toughness of ASTM A 533, Grade B, Class 1 steel was 349 kJ/m 2 and at the equivalent upper shelf temperature, the heat treated material exhibited 87 kJ/m 2 . The maximum load fracture toughness was found to be linearly proportional to specimen size, and only specimens which failed to meet ASTM size criteria exhibited maximum load toughness less than J/sub Ic/

  7. Potential high fluence response of pressure vessel internals constructed from austenitic stainless steels

    International Nuclear Information System (INIS)

    Garner, F.A.; Greenwood, L.R.; Harrod, D.L.

    1993-08-01

    Many of the in-core components in pressurized water reactors are constructed of austenitic stainless steels. The potential behavior of these components can be predicted using data on similar steels irradiated at much higher displacement rates in liquid-metal reactors or water-cooled mixed-spectrum reactors. Consideration of the differences between the pressurized water environment and that of the other reactors leads to the conclusion that significant amounts of void swelling, irradiation creep, and embrittlement will occur in some components, and that the level of damage per atomic displacement may be larger in the pressurized water environment

  8. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  9. Computational methods for fracture analysis of heavy-section steel technology (HSST) pressure vessel experiments

    International Nuclear Information System (INIS)

    Bass, B.R.; Bryan, R.H.; Bryson, J.W.; Merkle, J.G.

    1985-01-01

    This paper summarizes the capabilities and applications of the general-purpose and special-purpose computer programs that have been developed at ORNL for use in fracture mechanics analyses of HSST pressure vessel experiments. Emphasis is placed on the OCA/USA code, which is designed for analysis of pressurized-thermal-shock (PTS) conditions, and on the ORMGEN/ADINA/ORVIRT system which is used for more general analysis. Fundamental features of these programs are discussed, along wih applications to pressure vessel experiments. (orig./HP)

  10. Annealing of the RPV of unit 1 in Loviisa 1996

    International Nuclear Information System (INIS)

    Ahlstrand, R.; Kohopaeae, J.

    1997-01-01

    The critical circumferential core area weld of Loviisa 1 reactor pressure vessel was successfully annealed during the refueling and maintenance outage in August 1996. The weld was heated up to the annealing temperature of 475 deg.C and this temperature was maintained for 100 hours. The work was implemented by Skoda Nuclear Machinery Ltd as a main supplier representing consortium of Skoda Nuclear machinery Ltd from Czech Republic and Bohunice Nuclear Power Plant from Slovak Republic. Comprehensive material testing programs have been carried out to ensure the licensing of the annealing. Part of these programs have not yet been finished and are still going on. In the domestic programs sophisticated testing techniques including electric discharge machining and reconstitution techniques were used. Thus already tested surveillance specimens halves could be used as authentic material. The licensing work has been carried out mainly by VTT in Finland and Moht Otjig RM in Russia. A new comprehensive surveillance program has started to follow the embrittlement of the RPV after annealing. (author)

  11. Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M [AB Atomenergi, Nykoeping (Sweden); Myers, H P [Chalmers Institute of Technology, Goeteborg (Sweden); Hannerz, N E [Motala Verkstads AB, Motala (Sweden)

    1967-09-15

    Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10{sup 18} and 5.6 x 10{sup 18} n/cm{sup 2} (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature.

  12. Dislocation structures in 16MND5 pressure vessel steel strained in uniaxial tension

    Czech Academy of Sciences Publication Activity Database

    Obrtlík, Karel; Robertson, Ch.; Marini, B.

    2005-01-01

    Roč. 342, - (2005), s. 35-41 ISSN 0022-3115 R&D Projects: GA AV ČR(CZ) 1QS200410502 Institutional research plan: CEZ:AV0Z20410507 Keywords : bainitic steels * dislocation structure * low temperature deformation Subject RIV: JG - Metallurgy Impact factor: 1.414, year: 2005

  13. Irradiation Effects at 160-240 deg C in Some Swedish Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Grounes, M.; Myers, H.P.; Hannerz, N.E.

    1967-09-01

    Tensile specimens, Charpy impact specimens and miniature impact specimens of six steels in different conditions were irradiated to 2.8 x 10 18 and 5.6 x 10 18 n/cm 2 (> 1 MeV) at 160-240 deg C. The steels investigated were SIS 142103, 2103/R3, NO 345, Fortiweld, Fortiweld HS and OK 54 P. There is no correlation between the increase in transition temperature and initial transition temperature. However, changes in strength and ductility can be correlated to the initial yield strength. The increases in transition temperature due to strain aging and irradiation are approximately additive. The irradiation-induced changes in 2103/R3 and Fortiweld HS steels do not vary with position in the thickness of the plate. Different tempering treatments in Fortiweld HS steel do not change the extent of irradiation effects. Normal Charpy V-notch impact specimens and miniature specimens give the same irradiation-induced increase in transition temperature

  14. Hot Deformation Behavior of SA508Gr.4N Steel for Reactor Pressure Vessels

    Directory of Open Access Journals (Sweden)

    YANG Zhi-qiang

    2017-08-01

    Full Text Available The high-temperature plastic deformation and dynamic recrystallization behavior of SA508Gr.4N steel were investigated through hot deformation tests in a Gleeble1500D thermal mechanical simulator. The compression tests were performed in the temperature range of 1050-1250℃ and the strain rate range of 0.001-0.1s-1 with true strain of 0.16. The results show that from the high-temperature true stress-strain curves of the SA508Gr.4N steel, the main feature is dynamic recrystallization,and the peak stress increases with the decrease of deformation temperature or the increase of strain rate, indicating the experimental steel is temperature and strain rate sensitive material. The constitutive equation for SA508Gr.4N steel is established on the basis of the true stress-strain curves, and exhibits the characteristics of the high-temperature flow behavior quite well, while the activation energy of the steel is determined to be 383.862kJ/mol. Furthermore, an inflection point is found in the θ-σ curve, while the -dθ/dσ-σ curve shows a minimum value. The critical strain increases with increasing strain rate and decreasing deformation temperature. A linear relationship between critical strain (εc and peak strain (εp is found and could be expressed as εc/εp=0.517. The predicted model of critical strain could be described as εc=8.57×10-4Z0.148.

  15. Ultrasonic measurement on RPV stud-bolt loading under hot transient of Qinshan NPP

    International Nuclear Information System (INIS)

    Qu Jiadi; Dou Yikang; Zhu Shiming; Lu Jie; Wang Yingguan

    1994-01-01

    It is a continuation of research work for sealing analysis and tests on the PRV of PWR. It expounds that the key of solving thermal transient sealing problem lies in giving the thermal increment of stud-bolt fatigue life and transient loading spectrum for vessel analysis. The authors recounted the fundamental works and main results of ultrasonic measurement on RPV stud-bolt loading on the reactor of Qinshan Nuclear Power Plant. The measuring capability exceeds 1 m length and 300 degree C temperature. Therefore, it is possible to be used in the field of NPP

  16. Shallow crack effect on brittle fracture of RPV during pressurised thermal shock

    International Nuclear Information System (INIS)

    Ikonen, K.

    1995-12-01

    This report describes the study on behaviour of postulated shallow surface cracks in embrittled reactor pressure vessel subjected to pressurised thermal shock loading in an emergency core cooling. The study is related to the pressure vessel of a VVER-440 type reactor. Instead of a conventional fracture parameter like stress intensity factor or J integral the maximum principal stress distribution on a crack tip area is used as a fracture criteria. The postulated cracks locate circumferentially at the inner surface of the reactor pressure wall and they penetrate the cladding layer and open to the inner surface. Axisymmetric and semielliptical crack shapes were studied. Load is formed of an internal pressure acting also on crack faces and of a thermal gradient in the pressure vessel wall. Physical properties of material and loading data correspond real conditions in VVER-440 RPV. The study was carried out by making lot of 2D- and 3D- finite element calculations. Analysing principles and computer programs are explained. Except of studying the shallow crack effect, one objective of the study has also been to develop further expertise and the in-house developed computing system to make effectively elastic-plastic fracture mechanical analyses for real structures under complicated loads. Though the study concerns VVER-440 RPV, the results are of more general interest especially related to thermal loads. (orig.) (11 refs.)

  17. Comparative assessment of cyclic J-R curve determination by different methods in a pressure vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Chowdhury, Tamshuk, E-mail: tamshuk@gmail.com [Deep Sea Technologies, National Institute of Ocean Technology, Chennai, 600100 (India); Sivaprasad, S.; Bar, H.N.; Tarafder, S. [Fatigue & Fracture Group, Materials Science and Technology Division, CSIR-National Metallurgical Laboratory, Jamshedpur, 831007 (India); Bandyopadhyay, N.R. [School of Materials Science and Engineering, Indian Institute of Engineering, Science and Technology, Shibpur, Howrah, 711103 (India)

    2016-04-15

    Cyclic J-R behaviour of a reactor pressure vessel steel using different methods available in literature has been examined to identify the best suitable method for cyclic fracture problems. Crack opening point was determined by moving average method. The η factor was experimentally determined for cyclic loading conditions and found to be similar to that of ASTM value. Analyses showed that adopting a procedure analogous to the ASTM standard for monotonic fracture is reasonable for cyclic fracture problems, and makes the comparison to monotonic fracture results straightforward. - Highlights: • Different methods of cyclic J-R evaluation compared. • A moving average method for closure point proposed. • η factor for cyclic J experimentally validated. • Method 1 is easier, provides a lower bound and direct comparison to monotonic fracture.

  18. On the correlation between irradiation-induced microstructural features and the hardening of reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Lambrecht, M.; Meslin, E.; Malerba, L.; Hernandez-Mayoral, M.; Bergner, F.; Pareige, P.; Radiguet, B.; Almazouzi, A.

    2010-01-01

    A correlation is attempted between microstructural observations by various complementary techniques, which have been implemented within the PERFECT project and the hardening measured by tensile tests of reactor pressure vessel steel and model alloys after irradiation to a dose of ∼7 x 10 19 n cm -2 . This is done, using the simple hardening model embodied by the Orowan equation and applying the most suitable superposition law, as suggested by a parametric study using the DUPAIR line tension code. It is found that loops are very strong obstacles to dislocation motion, but due to their low concentration, they only play a minor role in the hardening itself. For the precipitates, the contrary is found, although they are quite soft (due to their very small sizes and their coherent nature), they still play the dominant role in the hardening. Vacancy clusters are important for the formation of both loops and precipitates, but they will play almost no role in the hardening by themselves.

  19. Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

    International Nuclear Information System (INIS)

    Kang, Han-Ok; Lee, Jun; Park, Cheon-Tae; Yoon, Ju-Hyeon; Park, Keun-Bae

    2007-01-01

    SMART is an advanced modular integral type pressurized water reactor for a seawater desalination and an electricity production. Major components of the reactor coolant system such as the pressurizer, Reactor Coolant Pump (RCP), and steam generators are located inside the reactor vessel. The SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs), improve the natural circulation capability, and better accommodate and thus enhance a resistance to a wide range of transients and accidents. The safety goals of the SMART are enhanced through highly reliable safety systems such as the passive residual heat removal system (PRHRS) and the safeguard vessel coupled with the passive safety injection feature. The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV, SIT, and the associated valves and pipelines. A primary function of the safeguard vessel is to confine any radioactive release from the primary circuit within the vessel under DBAs related to loss of the integrity of the primary system. A preliminary performance analysis program for a safety system using the safeguard vessel is developed in this study. The developed program is composed of several subroutines for the reactor coolant system, passive safety injection system, safeguard vessel including the pressure suppression pool, and PRHRS. A small break loss of coolant accident at the upper part of a reactor is analyzed and the results are discussed

  20. Interaction between molten corium UO{sub 2+x}-ZrO{sub 2}-FeO{sub y} and VVER vessel steel

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A. [Alexandrov Sci Res Technol Inst, Sosnovyi Bor (Russian Federation); Gusarov, V. V.; Almiashev, V. I. [Russian Acad Sci, Inst Silicate Chem, St Petersburg (Russian Federation); Lopukh, D. B. [SPb State Electrotech Univ LETI SPbGETU, St Petersburg (Russian Federation); Bottomley, D. [Joint Res Ctr, Inst Transurane, Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA Saclay, DEN, DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Dresden Rossendorf, Dresden (Germany); Fichot, F. [CEA Cadarache, SEMCA, DPAM, IRSN, St Paul Les Durance (France); Kymalainen, O. [FORTUM Nucl Serv Ltd, Espoo (Finland)

    2010-07-01

    In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO{sub 2+x}-ZrO{sub 2}-FeO{sub y} melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction Zone. The available experimental data have been used to develop a correlation for the corrosion rate as a function of temperature and heat flux. (authors)

  1. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Sun Mingyue, E-mail: mysun@imr.ac.cn [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China); Luhan, Hao; Shijian, Li; Dianzhong, Li; Yiyi, Li [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, 72 Wenhua Road, Shenyang 110016 (China)

    2011-11-15

    Highlights: > A series of flow stress constitutive equations for SA508-3 steel were successfully established. > The experimental results under different conditions have validated the constitutive equations. > An industrial application of the model was present to simulate a large conical shell forging process. - Abstract: Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  2. The role of phosphorus in the irradiation embrittlement of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Jones, R.B.; Buswell, J.T.

    1987-02-01

    An analysis has been performed of the influence of phosphorus on post-irradiation materials properties and microstructures determined on a variety of PWR steels and variants following exposure to MTR or reactor surveillance irradiations to doses not exceeding 7 x 10 19 n.cm -2 (E>1.0MeV) at 250-290 0 C. The irradiation-induced shifts in impact transition temperature, matrix hardening and the relative small angle neutron scattering response were found to rise most rapidly with increasing phosphorus when the copper content of the steel was 0.03 w/o. The sensitivity of the changes in mechanical properties to phosphorus content decreased as the copper content was increased. At copper levels typical of modern PWR steel manufacture (Cu 3 P) produced by the irradiation induced segregation of phosphorus to defect sinks and the depletion of phosphorus in solid solution as detected by high sensitivity electron microscopy and other analytical techniques. At higher levels of copper (approx. 0.3 w/o) the effect of phosphorus on properties was reduced by a factor of three due to the observed incorporation of phosphorus into the small copper precipitates formed during irradiation. Grain boundary embrittlement by phosphorus under irradiation is not thought to be important but further evidence concerning the post-irradiation fracture mode and the development of the deleterious influence of phosphorus with irradiation dose is required for a comprehensive understanding of its action. Some suggestions for future work are made. (author)

  3. Modeling flow stress constitutive behavior of SA508-3 steel for nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Sun Mingyue; Hao Luhan; Li Shijian; Li Dianzhong; Li Yiyi

    2011-01-01

    Highlights: → A series of flow stress constitutive equations for SA508-3 steel were successfully established. → The experimental results under different conditions have validated the constitutive equations. → An industrial application of the model was present to simulate a large conical shell forging process. - Abstract: Based on the measured stress-strain curves under different temperatures and strain rates, a series of flow stress constitutive equations for SA508-3 steel were firstly established through the classical theories on work hardening and softening. The comparison between the experimental and modeling results has confirmed that the established constitutive equations can correctly describe the mechanical responses and microstructural evolutions of the steel under various hot deformation conditions. We further represented a successful industrial application of this model to simulate a forging process for a large conical shell used in a nuclear steam generator, which evidences its practical and promising perspective of our model with an aim of widely promoting the hot plasticity processing for heavy nuclear components of fission reactors.

  4. Development of a crack monitoring technique for use in a corrosion fatigue study of SA533-B pressure vessel steel

    International Nuclear Information System (INIS)

    Benson, J.M.; Tait, R.B.; Garrett, G.G.

    1981-10-01

    At present there does not exist a realistic crack growth law which will provide a good description of the relationship between the alternating stress intensity factor and the crack growth per cycle of stress. Such a law should be applicable to either the pressurized water reactor environment (PWR) or boiling water reactor environmnt (BWR). This project was formulated with the aim of examining the fatigue crack growth rate of SA533-B steel (a nuclear pressure vessel steel) in the threshold region in a simulated PWR environment. The aim of this report is to develop a