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.

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.

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

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

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

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

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)

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

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.)

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

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.

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

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.

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)

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

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

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

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.

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

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

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

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

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.

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.

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.

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

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)

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

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.

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)

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

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)

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

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

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

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

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)

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)

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.

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

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.

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.

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)

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)

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)

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)

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.

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)

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.

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)

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.

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

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

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)

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

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)

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.

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

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.

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.

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

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

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

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)

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

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

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)

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.

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.)

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).

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.

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

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)

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

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

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.)

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

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.

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.

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

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

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)

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.)

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.

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)

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

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)

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)

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

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

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)

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

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

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)

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)

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

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.

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

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

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

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

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.

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.

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

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.

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)

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)

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.

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.

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.

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

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

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

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

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

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

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.

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.

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

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.

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

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.

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

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.

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.

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.)

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

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

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

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

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

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

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

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

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.

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

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

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)

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

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

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

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

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

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.

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.

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

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)

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)

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.

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)

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

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

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

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)

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

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)

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)

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

Resolution of the Task A-11 reactor-vessel materials-toughness safety issue. Appendices C-K

International Nuclear Information System (INIS)

1982-10-01

The central problem in the unresolved safety issue A-11, Reactor Vessel Materials Toughness, was to provide guidance in performing analyses required by 10 CFR Part 50, Appendix G, Section V.C. for reactor pressure vessels (RPVs) which fail to meet the toughness requirement during service life as a result of neutron radiation embrittlement. Although the methods of linear-elastic fracture mechanics (LEFM) were adequate for low-temperature RPV problems, they were inapplicable under operating conditions because vessel steels, even those which exhibit less than 50 ft-lb of C/sub v/ energy, were relatively tough at temperatures where the impact energy reached its upper shelf values. A technical team of recognized experts was organized to assist the NRC staff in addressing the problem. Using the foundation of the tearing modulus concept, which had been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the problem of RPV fracture with assumed beltline region flaws. The first paper of this report is a summary of the problem, the solutions, and the results of verification analyses. The details are provided in a series of appendices in Volumes I and II

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...

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

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)

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)

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

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

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.

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.)

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

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.)

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.)

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.

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)

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)

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

Development of automatic reactor vessel inspection systems; development of data acquisition and analysis system for the nuclear vessel weld

Energy Technology Data Exchange (ETDEWEB)

Lee, Jong Po; Park, C. H.; Kim, H. T.; Noh, H. C.; Lee, J. M.; Kim, C. K.; Um, B. G. [Research Institute of KAITEC, Seoul (Korea)

2002-03-01

The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine heavy vessel welds. In order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet. In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed. In this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition software was developed. The new systems were tested on the RPV welds of Ulchin Unit 6 to confirm their functions and capabilities. They worked very well as designed and the tests were successfully completed. 13 refs., 34 figs., 11 tabs. (Author)

Comparison of rate theory based modeling calculations with the surveillance test results of Korean light water reactors

International Nuclear Information System (INIS)

Lee, Gyeong Geun; Lee, Yong Bok; Kim, Min Chul; Kwon, Junh Yun

2012-01-01

Neutron irradiation to reactor pressure vessel (RPV) steels causes a decrease in fracture toughness and an increase in yield strength while in service. It is generally accepted that the growth of point defect cluster (PDC) and copper rich precipitate (CRP) affects radiation hardening of RPV steels. A number of models have been proposed to account for the embrittlement of RPV steels. The rate theory based modeling mathematically described the evolution of radiation induced microstructures of ferritic steels under neutron irradiation. In this work, we compared the rate theory based modeling calculation with the surveillance test results of Korean Light Water Reactors (LWRs)

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.)

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

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.

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)

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)

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

International Cooperation for the Dismantling of Chooz A Reactor Pressure Vessel

International Nuclear Information System (INIS)

Grenouillet, J.J.; Posivak, E.

2009-01-01

Chooz A is the first PWR that is being decommissioned in France. The main issue that is conditioning the success of the project is the Reactor Pressure Vessel (RPV) and Reactor Vessel Internals (RVI) segmentation. Whereas Chooz A is the first and unique RPV and RVI being dismantled in France, there are many similar experiences available in the world. Thus the project team was eager to cooperate with other teams facing or being faced with the same issue. A cooperation programme was established in two separate ways: - Benefiting from experience feedback from completed RPV and RVI dismantling projects, - Looking for synergy with future RPV dismantling projects for activities such as segmentation tools design, qualification and manufacturing for example. This paper describes the implementation of this programme and how the outcome of the cooperation was used for the implementation of Chooz-A RPV and RVI segmentation project. It shows also the limits of such a cooperation. (authors)

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)

In-Vessel Retention via External Reactor Cooling

Energy Technology Data Exchange (ETDEWEB)

Bachrata, Andrea [CTU in Prague, Faculty of nuclear sciences and physical engineering, V Holesovickach 2 180 00, Prague 8 (Czech republic)

2008-07-01

In-vessel (corium) retention (IVR) via external reactor pressure vessel (RPV) cooling is considered to be an effective severe accident management strategy for corium localisation and stabilisation. The main idea of IVR strategy consists in flooding the reactor cavity and transferring the decay heat through the wall of RPV to the recirculating water and than to the atmosphere of the containment of nuclear power plant. The aim of this strategy is to localise and to stabilise the corium inside the RPV. Not using this procedure could destroy the integrity of RPV and might cause the interaction of the corium with the concrete at the bed of the reactor cavity. Several experimental facilities and computer codes (MVITA, ASTEC module DIVA and CFD codes) were applied to simulate the IVR strategy for concrete reactor designs. The necessary technical modifications concerning the implementation of IVR concept were applied at the Loviisa NPP (VVER-440/V213). This strategy is also an important part of the advanced reactor designs AP600 and AP1000. (authors)

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)

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

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)

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)

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)

Evaluation of microstructural difference in low alloy steel (SA508 CI.3) by magnetic measurements

International Nuclear Information System (INIS)

Chi, Se Hwan; Park, Duck Gun; Hong, Jun Hwa; Kuk, Il Hium; Song, Sook Hyang; Shur, Dong Soo

1997-01-01

Magnetic measurements on SA 508 Cl. 3 low alloy reactor pressure vessel (RPV) steels of three different refining processes were conducted to prepare baseline magnetic properties for future comparison with that of irradiated one and to examine the correlation between microstructures and magnetic parameters. TEM(transmission electron microscopy) on thin films and carbon replications were performed for microstructural investigation, and microhardness, tensile and fracture toughness tests were conducted for mechanical properties measurement. Barkhausen noise analysis was conducted and hysteresis parameters were measured. Microstructure, mechanist and magnetic properties were interpreted together to explore their inter-relationship. Also the applicability of a magnetic nondestructive evaluation method for monitoring fine microstructural changes in low alloy steels of complicated microstructure was examined. The results showed that, of all magnetic parameters, Barkhausen noise amplitudes show consistent change with microstructure (grain size, carbide microstructure, lath width) and hardness of the materials. For monitoring microstructural and mechanical property changes of low alloy steel under service environment of reactor pressure vessel steels, Barkhausen noise amplitude and coercivity appear to be as key magnetic parameters useful for nondestructive evaluation.

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)

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

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

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

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

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.

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.)

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)

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.)

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.

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

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)

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

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

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

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

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.

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.)

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

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

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.

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)

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)

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

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

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.

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.

Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

International Nuclear Information System (INIS)

Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

2017-01-01

Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests

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)

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)

The role of point defect clusters in reactor pressure vessel embrittlement

International Nuclear Information System (INIS)

Stoller, R.E.

1993-01-01

Radiation-induced point defect clusters (PDC) are a plausible source of matrix hardening in reactor pressure vessel (RPV) steels in addition to copper-rich precipitates. These PDCs can be of either interstitial or vacancy type, and could exist in either 2 or 3-D shapes, e.g. small loops, voids, or stacking fault tetrahedra. Formation and evolution of PDCs are primarily determined by displacement damage rate and irradiation temperature. There is experimental evidence that size distributions of these clusters are also influenced by impurities such as copper. A theoretical model has been developed to investigate potential role of PDCs in RPV embrittlement. The model includes a detailed description of interstitial cluster population; vacancy clusters are treated in a more approximate fashion. The model has been used to examine a broad range of irradiation and material parameters. Results indicate that magnitude of hardening increment due to these clusters can be comparable to that attributed to copper precipitates. Both interstitial and vacancy type defects contribute to this hardening, with their relative importance determined by the specific irradiation conditions

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)

Consequence evaluation of hypothetical reactor pressure vessel support failure

International Nuclear Information System (INIS)

Lu, S.C.; Holman, G.S.; Lambert, H.E.

1991-01-01

This paper describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. The structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports and that the SG supports and the RCP supports have sufficient design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas for further investigation and concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns. (author)

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

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

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

Estimation of radiation hardening in ferritic steels using the cluster dynamics models

Energy Technology Data Exchange (ETDEWEB)

Kwon, Jun Hyun; Kim, Whung Whoe; Hong, Jun Hwa [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

Evolution of microstructure under irradiation brings about the mechanical property changes of materials, of which the major concern is radiation hardening in this work. Radiation hardening is generally expressed in terms of an increase in yield strength as a function of radiation dose and temperature. Cluster dynamics model for radiation hardening has been developed to describe the evolution of point defects clusters (PDCs) and copperrich precipitates (CRPs). While the mathematical models developed by Stoller focus on the evolution of PDCs in ferritic steels under neutron irradiation, we slightly modify the model by including the CRP growth and estimate the magnitude of hardening induced by PDC and CRP. The model is then used to calculate the changes in yield strength of RPV steels. The calculation results are compared to measured yield strength values, obtained from surveillance testing of PWR vessel steels in France.

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

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

1999-01-01

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

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

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.

Materials for high temperature reactor vessels

International Nuclear Information System (INIS)

Buenaventura Pouyfaucon, A.

2004-01-01

Within the 5th Euraton Framework Programme, a big effort is being made to promote and consolidate the development of the High Temperature Reactor (HTR). Empresarios Agrupados is participating in this project and among others, also forms part of the HTR-M project Materials for HTRs. This paper summarises the work carried out by Empresarios Agrupados regarding the material selection of the HTR Reactor Pressure Vessel (RPV). The possible candidate materials and the most promising ones are discussed. Design aspects such as the RPV sensitive zones and material damage mechanisms are considered. Finally, the applicability of the existing design Codes and Standards for the design of the HTR RPV is also discussed. (Author)

Optimization of reactor pressure vessel internals segmentation in Korea

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-15

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

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)

Assessment of In-vessel corium retention for VVER-440/V213

International Nuclear Information System (INIS)

Matejovic, P.; Barnak, M.; Bachraty, M.; Berky, R.

2011-01-01

In-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising severe accident management strategy for VVER-440/V213 reactors. In general, the avoiding of boiling crisis on outer (cooled) RPV (reactor pressure vessel) surface is sufficient condition for preserving the RPV integrity. The crucial point of the proposed IVR concept for VVER-440/V213 is the narrow gap between elliptical lower head and thermal and biological shield. In the cold conditions the width of this gap is only about 2 cm and would be even lower in hot IVR conditions, when the reactor wall is subjected to large thermal gradients due to temperature difference between the hot inner surface (loaded by corium) and cold outer surface (which is cooled by water in flooded cavity). Sufficient gap should remain free for coolant flow for the success of the proposed IVR concept. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width are of primarily importance. Two different approaches were used for the estimation of the thermal load: a conservative approach and a transient approach, both were computed with the ASTEC code. The structural analysis of RPV subjected to IVR load was performed using the finite element method (FEM) code ANSYS release 10.0. From the results obtained it follows, that even when the RPV is subjected to limiting loading conditions during severe accident, there should be sufficient gap width (∼ 1 cm) between RPV wall and thermal/biological shield for the coolant flow in natural circulation regime alongside the outer surface of the RPV wall

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

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)

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)

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.)

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.

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

Coupled thermo-mechanical analysis of corium-loaded lower head of pressure vessel

International Nuclear Information System (INIS)

Mishra, J.; Balasubramaniyan, V.

2016-01-01

A severe accident in the pressurised water reactor may lead to the relocation of core materials to the lower head of Reactor Pressure Vessel (RPV). The core debris at the bottom of RPV forms a melt pool of corium due to decay heat. The understanding of behaviour of pressure vessel, characterised by failure mode and time to failure, in this scenario is one of the important steps in predicting the accident progression. The most predominant failure mode is multi-axial creep deformation of the vessel with a non-uniform temperature field. Towards this, a numerical analysis methodology is developed for the prediction of pressure vessel deformation during the severe accidents. The methodology involves 2-D finite element modelling under multi-physics environment, which account the creep phenomena using Norton-Bailey creep law with a typical damage model of RPV material. The validation of the methodology is carried out using the results from OLHF experiment carried out in Sandia National Laboratory (SNL), USA, within the framework of an OECD. (author)

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.

Development of automatic reactor vessel inspection systems: development of data acquisition and analysis system for the nuclear vessel weld

Energy Technology Data Exchange (ETDEWEB)

Park, C. H.; Lim, H. T.; Um, B. G. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

2001-03-01

The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine the reactor vessel weldsIn order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed in this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition and analysis software was developed. 11 refs., 6 figs., 9 tabs. (Author)

RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

International Nuclear Information System (INIS)

Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.

1984-01-01

Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

Multivariable modeling of pressure vessel and piping J-R data

International Nuclear Information System (INIS)

Eason, E.D.; Wright, J.E.; Nelson, E.E.

1991-05-01

Multivariable models were developed for predicting J-R curves from available data, such as material chemistry, radiation exposure, temperature, and Charpy V-notch energy. The present work involved collection of public test data, application of advanced pattern recognition tools, and calibration of improved multivariable models. Separate models were fitted for different material groups, including RPV welds, Linde 80 welds, RPV base metals, piping welds, piping base metals, and the combined database. Three different types of models were developed, involving different combinations of variables that might be available for applications: a Charpy model, a preirradiation Charpy model, and a copper-fluence model. In general, the best results were obtained with the preirradiation Charpy model. The copper-fluence model is recommended only if Charpy data are unavailable, and then only for Linde 80 welds. Relatively good fits were obtained, capable of predicting the values of J for pressure vessel steels to with a standard deviation of 13--18% over the range of test data. The models were qualified for predictive purposes by demonstrating their ability to predict validation data not used for fitting. 20 refs., 45 figs., 16 tabs

Feasibility study for CPR1000 incore measurement instrumentation educed from the reactor pressure vessel upper head

International Nuclear Information System (INIS)

Guang Jianwei; Liu Qian; Li Wenhong; Duan Yuangang

2010-01-01

The article discusses about the feasibility of in-core measurement instrumentation educed from the reactor pressure vessel (RPV) upper head. Incore instrumentation educed from the reactor pressure vessel upper head is one of advanced technology in the third generation nuclear power plant. This technology can reduce the manufacture problem of RPV; decrease the manufacture time effectively. Furthermore, this technology can get rid of the trouble for loss of water caused by many penetrations in the RPV bottom head, can increase security of nuclear power plant. By the description of structure analysis, comparison, maturity for four type incore instrumentation detectors, the incore instrumentation can be educed from RPV upper head, which can increase reactor's security, reduce the manufacture time, decrease group dose in refueling period. The core design ability can be enhanced through this study. (authors)

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)

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

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)

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.)

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

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

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

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.)

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).

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

Results of reactor pressure vessels ISI

International Nuclear Information System (INIS)

Cepcek, S.

1994-01-01

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

Guidelines for prediction of irradiation embrittlement of operating WWER-440 reactor pressure vessels

International Nuclear Information System (INIS)

2005-06-01

This TECDOC has been developed under an International Atomic Energy Agency Coordinated Research Project (CRP) entitled Evaluation of Radiation Damage of WWER Reactor Pressure Vessels (RPV) using Database on RPV Materials to develop the guidelines for prediction of radiation damage to WWER-440 PRVs. The WWER-440 RPV was designed by OKB Gidropress, Russian Federation, the general designer. Prediction of irradiation embrittlement of RPV materials is usually done in accordance with relevant codes and standards that are based on the large amounts of information from surveillance and research programmes. The existing Russian code (standard for strength calculations of components and piping in NPPs - PNAE G 7-002-86) for the WWER RPV irradiation embrittlement assessment was approved more than twenty years ago and based mostly on the experimental data obtained in research reactors with accelerated irradiation. Nevertheless, it is still in use and generally consistent with new data. The present publication presents the analyses using all available data required for more precise prediction of radiation embrittlement of WWER-440 RPV materials. Based on the fact that it contains a large amount of data from surveillance programmes as well as research programmes, the IAEA International Database on RPV Materials (IDRPVM) is used for the detailed analysis of irradiation embrittlement of WWER RPV materials. Using IDRPVM, the guideline is developed for assessment of irradiation embrittlement of RPV ferritic materials as a result of degradation during operation. Two approaches, i.e. transition temperatures based on Charpy impact notch toughness, as well as based on static fracture toughness tests, are used in RPV integrity evaluation. The objectives of the TECDOC are the analysis of irradiation embrittlement data for WWER- 440 RPV materials using IDRPVM database, evaluation of predictive formulae depending on chemical composition of the material, neutron fluence, flux, and

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

A study on the sealing performance of metallic C-rings in reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Jia, Xiaohong, E-mail: jiaxh@mail.tsinghua.edu.cn [State Key Laboratory of Tribology, Tsinghua University, Beijing 100084 (China); Chen, Huaming [State Key Laboratory of Tribology, Tsinghua University, Beijing 100084 (China); Li, Xinggen [Ningbo Tiansheng Sealing Packing Co., Ltd, Ningbo 315302 (China); Wang, Yuming [State Key Laboratory of Tribology, Tsinghua University, Beijing 100084 (China); Wang, Longke [Eaton Corporation, MN (United States)

2014-10-15

Highlights: • FE analysis on compression–resilience of metallic C-ring is performed and validated by experiments. • Model of RPV sealing system including the C-rings is developed. • Deformation data from factory hydraulic test of the RPV are used to verify the model. • C-rings’ behavior under designing condition is analyzed. • The model provides a reliable evaluation on the sealing performance of RPV. - Abstract: Double metallic C-rings are used in pressure vessel of pressurized water reactor (PWR) to seal the bolt-connected flanges. To evaluate the sealing performance, it is necessary to study both the C-rings’ intrinsic properties and their behavior in reactor pressure vessel (RPV) under various loading conditions. The compression–resilience property and linear load are the basic information to evaluate the performance of a well-designed C-ring's. An equivalent model of C-ring is constructed by means of ANSYS to analyze its intrinsic properties, and is also validated by experiments on scaled samples. This model is applied to develop a 2D-axisymmetric FE model of sealing system including RPV and C-rings with the consideration of nonlinear material, contacting problem and multiple coupled effects. The simulation results of RPV deformation under the hydraulic test condition agree well with the data of factory hydraulic test. With the verified model, an analysis under the designing condition is performed to study C-rings’ behavior in the RPV, and then provides a reliable evaluation on the sealing performance of RPV.

PWR reactor vessel in-service-inspection according to RSEM

International Nuclear Information System (INIS)

Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

2006-01-01

Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

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.

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)

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

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.

Proposal of In-vessel corium retention concept for Paks NPP

International Nuclear Information System (INIS)

Elter, J.; Toth, E.; Matejovic, P.

2011-01-01

The in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) seems to be a promising severe accident management strategy not only for new generation of advanced PWRs, but also for VVER-440/V213 reactors, which were designed several years ago. The basic idea of in-vessel retention of corium is to prevent RPV failure by flooding the reactor cavity so that the reactor pressure vessel is submerged in water up to its support structures, and thus the decay heat can be transferred from the corium pool through the vessel wall and into the water surrounding the vessel. An IVR concept with simple ECVR loop based only on minor modifications of existing plant technology was proposed for the Paks Nuclear Power Plant. 2 severe accident (LB and SB LOCA) without availability of HP and LP safety injection in power upgrade (108%) conditions were simulated using the ASTEC code. The analyses show that the proposed solution is effective in preserving RPV integrity in the case of severe accident. Possible uncertainties in code predictions are covered by the applied conservative assumptions

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)

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)

Microstructural Characterization of Clad Interface in Welds of Ni-Cr-Mo High Strength Low Alloy Steel

Energy Technology Data Exchange (ETDEWEB)

Kim, Hong-Eun; Kim, Min-Chul; Lee, Ho-Jin; Kim, Keong-Ho [KAERI, Daejeon (Korea, Republic of); Lee, Ki-Hyoung [KAIST, Daejeon (Korea, Republic of); Lee, Chang-Hee [Hanyang Univ., Seoul (Korea, Republic of)

2011-08-15

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at 610°C for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

Microstructural Characterization of Clad Interface in Welds of Ni-Cr-Mo High Strength Low Alloy Steel

International Nuclear Information System (INIS)

Kim, Hong-Eun; Kim, Min-Chul; Lee, Ho-Jin; Kim, Keong-Ho; Lee, Ki-Hyoung; Lee, Chang-Hee

2011-01-01

SA508 Gr.4N Ni-Cr-Mo low alloy steel, in which Ni and Cr contents are higher than in commercial SA508 Gr.3 Mn-Mo-Ni low alloy steels, may be a candidate reactor pressure vessel (RPV) material with higher strength and toughness from its tempered martensitic microstructure. The inner surface of the RPV is weld-cladded with stainless steels to prevent corrosion. The goal of this study is to evaluate the microstructural properties of the clad interface between Ni-Cr-Mo low alloy steel and stainless weldment, and the effects of post weld heat treatment (PWHT) on the properties. The properties of the clad interface were compared with those of commercial Mn-Mo-Ni low alloy steel. Multi-layer welding of model alloys with ER308L and ER309L stainless steel by the SAW method was performed, and then PWHT was conducted at 610°C for 30 h. The microstructural changes of the clad interface were analyzed using OM, SEM and TEM, and micro-Vickers hardness tests were performed. Before PWHT, the heat affected zone (HAZ) showed higher hardness than base and weld metals due to formation of martensite after welding in both steels. In addition, the hardness of the HAZ in Ni-Cr-Mo low alloy steel was higher than that in Mn-Mo-Ni low alloy steel due to a comparatively high martensite fraction. The hardness of the HAZ decreased after PWHT in both steels, but the dark region was formed near the fusion line in which the hardness was locally high. In the case of Mn-Mo-Ni low alloy steel, formation of fine Cr-carbides in the weld region near the fusion line by diffusion of C from the base metal resulted in locally high hardness in the dark region. However, the precipitates of the region in the Ni-Cr-Mo low alloy steel were similar to that in the base metal, and the hardness in the region was not greatly different from that in the base metal.

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.)

Consequence evaluation of radiation embrittlement of Trojan reactor pressure vessel supports

International Nuclear Information System (INIS)

Lu, S.C.; Sommer, S.C.; Johnson, G.L.; Lambert, H.E.

1990-10-01

This report describes a consequence evaluation to address safety concerns raised by the radiation embrittlement of the reactor pressure vessel (RPV) supports for the Trojan nuclear power plant. The study comprises a structural evaluation and an effects evaluation and assumes that all four reactor vessel supports have completely lost the load carrying capability. By demonstrating that the ASME code requirements governing Level D service limits are satisfied, the structural evaluation concludes that the Trojan reactor coolant loop (RCL) piping is capable of transferring loads to the steam generator (SG) supports and the reactor coolant pump (RCP) supports. A subsequent design margins to accommodate additional loads transferred to them through the RCL piping. The effects evaluation, employing a systems analysis approach, investigates initiating events and the reliability of the engineered safeguard systems as the RPV is subject to movements caused by the RPV support failure. The evaluation identifies a number of areas of additional safety concerns, but further investigation of the above safety concerns, however, concludes that a hypothetical failure of the Trojan RPV supports due to radiation embrittlement will not result in consequences of significant safety concerns

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

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)

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

Investigation of Ductile-to-Brittle Transition of RPV Materials by using the Pre-cracked Charpy Impact Data

International Nuclear Information System (INIS)

Lee, Ho Jin; Lee, Bong Sang; Hong, Jun Hwa

2005-01-01

Much recent work in the field of elastic-plastic fracture mechanics has been directed to developing a mechanics-based relationship between the onset of cleavage fracture in structural components and that of Charpy V-notch specimens. The assessing processes of the cracks located in the reactor pressure vessel (RPV) is described in the ASME code Sec. III, App. G and Sec. XI, App. A. The RTNDT obtained from the impact test using standard Charpy V-notch (CVN) specimens is used as a reference temperature to assess the integrity of RPV materials. The initial RTNDT, for the Linde 80 weld, was determined by the 67.8J Charpy impact energy instead of drop weight test. Generally, Linde 80 weld has low upper-shelf energy. The initial RTNDT obtained from the Charpy impact energy curve has been considered overly conservative. Recently, master curve method has been investigated to assess the integrity of RPV materials directly. The initial RTT0 obtained from the master curve method is considered more realistic than the initial RTNDT obtained from impact test for low upper-shelf fracture toughness RPV materials. In this research, the correlation of transition regions between the master curves and the Charpy impact energy curves was investigated using the dynamic fracture toughness curve and the impact energy curve obtained from the impact test of pre-cracked Charpy (PCC) specimens. For the low toughness RPV material the ductile-to-brittle transition corresponding to the static master curve was anticipated using the invested correlation

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.

Reactor pressure vessel. Status report

International Nuclear Information System (INIS)

Elliot, B.J.; Hackett, E.M.; Lee, A.D.

1996-10-01

This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff's reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date

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

An effective surveillance strategy for reactor pressure vessel assessment in the long term operation perspective

International Nuclear Information System (INIS)

Chaouadi, R.; Gerard, R.

2015-01-01

The reactor pressure vessel (RPV) irradiation embrittlement is monitored by means of surveillance capsules containing the RPV belt-line materials, inserted inside the reactor pressure vessel (RPV) before the start of operation. These capsules are placed at location where they receive a higher neutron flux than the vessel wall, by a factor of the order of 2 to 3. They are regularly retrieved and tested to evaluate the RPV irradiation embrittlement according to specific regulatory procedures and standards, in order to guarantee the safe operation of the RPV throughout its lifetime. These procedures are often relying on empirical but conservative concepts. In parallel, material research reactor (MTR) irradiations are often used to support the surveillance data and to develop a better understanding of irradiation effects, not only qualitatively but also quantitatively. Taking advantage of the increased understanding of irradiation effects, analytical tools were developed to improve the evaluation embrittlement and quality assurance of the RPV embrittlement assessment. In this framework, an alternative but complementary surveillance program assessment was developed in Belgium, the so-called enhanced surveillance, in order to benefit from the latest developments in the area of materials science and irradiation effects. The neutron flux and fracture properties of the surveillance materials can be reliably characterized and correlated to each other using physically-based rather than empirical concepts. The enhanced surveillance approach is complementary to the mandatory regulatory procedure and allows quantifying the conservatism of the regulatory approach. The enhanced surveillance approach that uses the reconstitution technology to fabricate additional small size specimens, appropriate modeling tools and microstructural examination when required, makes it possible to rationalize all available information in a physically-based way

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

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.

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

Seal analysis technology for reactor pressure vessel

International Nuclear Information System (INIS)

Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

2009-01-01

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

The utility industry and reactor surveillance

International Nuclear Information System (INIS)

Jenkins, R.B.

1983-01-01

Every commercial nuclear power reactor pressure vessel (RPV) is required to have a reactor vessel surveillance program at the time of plant licensing. The program is part of a continuing structural integrity assessment of the RPV. As such, the surveillance program supplements Section III of the American Society of Mechanical Engineers (ASME) Code (1), which is the design basis for nuclear power plant component pressure boundaries. The Code assumes that the materials of construction are ductile in the evaluation and design of all components. The surveillance program for each RPV is intended to provide assurance of continued applicability of the ASME Code, Appendix G, assessment of that RPV's operating limits. This assessment ensures that the RPV is always in a condition which precludes the unstable propagation of flaws in the vessel wall material. The potential presence of flaws and the desire to ensure ductility are significant considerations in ferritic steels such as those used to fabricate nuclear reactor pressure vessels. These materials are known to exhibit transition from ductile-to-brittle fracture behavior over a determined temperature range. Neutron irradiation tends to shift this ductile-to-brittle behavior transition zone to a temperature higher than unirradiated materials

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

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)

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.

Regulatory Experience on Structural Integrity Issues of The Oldest Reactor Pressure Vessel in Korea

International Nuclear Information System (INIS)

Lee, Sang-Min; Cho, Doo-Ho; Kim, Jin-Su; Kim, Yong-Beum; Chung, Hae-Dong; Kim, Se-Chang; Choi, Jae-Boong

2015-01-01

A reactor pressure vessel plays a crucial role of retaining reactor coolant and core assemblies. The RPV integrity should be evaluated in consideration with the design transient condition and the material deterioration of RPV belt-line region. Especially, the pressurized thermal shock has been considered as one of the most important issues regarding the RPV integrity since Rancho Seco nuclear power plant accident in 1978. In this paper, the structural integrity evaluation of the oldest RPV in Korea was performed by using finite element analysis. PTS conditions like small break loss of coolant accident and Turkey Point steam line break were applied as loading conditions. Neutron fluence data equivalent to 40 years was used to determine the fracture toughness of RPV material. The 3-dimensional finite element model including a circumferential surface flaw was considered for fracture mechanics analysis. The RPV integrity was evaluated according to Japan Electric Association Code. (authors)

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.

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.

Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

International Nuclear Information System (INIS)

Server, W.L.; Nanstad, R.K.; Odette, G.R.

2012-01-01

The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

Proceedings of the Workshop on in-vessel core debris retention and coolability

International Nuclear Information System (INIS)

1999-01-01

This conference on in-vessel core debris retention and coolability is composed of 37 papers grouped in three sessions: session 1 (Keynote papers: Key phenomena of late phase core melt progression, accident management strategies and status quo of severe fuel damage codes, In-vessel retention as a severe accident management scheme, GAREC analyses in support of in-vessel retention concept, Latest findings of RASPLAV project); session 2 - Experiments and model development with five sub-sessions: sub-session 1 (Debris bed heat transfer: Debris and Pool Formation/Heat Transfer in FARO-LWR: Experiments and Analyses, Evaporation and Flow of Coolant at the Bottom of a Particle-Bed modelling Relocated Debris, Investigations on the Coolability of Debris in the Lower Head with WABE-2D and MESOCO-2D, Uncertainty and Sensitivity Analysis of the Heat Transfer Mechanisms in the Lower Head, Simulation of the Arrival and Evolution of Debris in a PWR Lower Head with the SFD ICARE2 code), sub-session 2 (Corium properties, molten pool natural convection, and crust formation: Physico-chemistry and corium properties for in-vessel retention, Experimental data on heat flux distribution from volumetrically heated pool with frozen boundaries, Thermal hydraulic phenomena in corium pools - numerical simulation with TOLBIAC and experimental validation with BALI, TOLBIAC code simulations of some molten salt RASPLAV experiments, SIMECO experiments on in-vessel melt pool formation and heat transfer with and without a metallic layer, Numerical investigation of turbulent natural convection heat transfer in an internally-heated melt pool and metallic layer, Current status and validation of CON2D and 3D code, Free convection of heat-generating fluid in a constrained during experimental simulation of heat transfer in slice geometry), sub-session 3 (Gap formation and gap cooling: Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water, Experimental investigations

Irradiation, annealing, and reirradiation research in the ORNL heavy-section steel irradiation program

International Nuclear Information System (INIS)

Nanstad, R.K.; Iskander, S.K.; McCabe, D.E.; 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 experimental results from work performed as part of the Heavy-Section Steel Irradiation (HSSI) Program managed by Oak Ridge National Laboratory (ORNL) for the U.S. Nuclear Regulatory Commission. The HSSI Program focuses on annealing and re-embrittlement response of materials which are representative of those in commercial RPVs and which are considered to be radiation-sensitive. Experimental studies include (1) the annealing of materials in the existing inventory of previously irradiated materials, (2) reirradiation of previously irradiated/annealed materials in a collaborative program with the University of California, Santa Barbara (UCSB), (3) irradiation/annealing/reirradiation of U.S. and Russian materials in a cooperative program with the Russian Research Center-Kurchatov Institute (RRC-KI), (4) the design and fabrication of an irradiation/anneal/reirradiation capsule and facility for operation at the University of Michigan Ford Reactor, (5) the investigation of potential for irradiation-and/or thermal-induced temper embrittlement in heat-affected zones (HAZs) of RPV steels due to phosphorous segregation at grain boundaries, and (6) investigation of the relationship between Charpy impact toughness and fracture toughness under all conditions of irradiation, annealing, and reirradiation

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.)

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.

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.)

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

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.

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

Development of data acquisition and analysis system for the nuclear vessel weld

Energy Technology Data Exchange (ETDEWEB)

Lee, J. P.; Park, C. H.; Lim, H. T.; Noh, H. C. [Research Institute of KAITEC, Taejeon (Korea)

2000-03-01

The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine heavy vessel welds. In order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet. In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed. In this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition software was developed. 11 refs., 6 figs. (Author)

Inspection accessibility study of the Millstones Unit 1 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Calhoun, G.; Kapoor, A.; Davis, J.B.

1990-07-01

The need for more extensive ultrasonic examination of the Reactor Pressure Vessel (RPV) shell welds at some of the BWR plants is becoming apparent. Recent NRC comments regarding limits on future inspection relief requests and proposed revisions to the ASME Boiler and Pressure Vessel Code, Section 11 have caused BWR plant owners and vendors of inspection services to look at inspection tooling that would allow for the ultrasonic examination of RPV shell welds previously considered inaccessible. One approach to inspection of the RPV shell welds that will allow for greater coverage at some BWR plants is to perform these inspections from inside the vessel. In late 1988, Westinghouse received a contract from Northeast Utilities to examine the flange ligament areas and the shell welds in the upper portion of the Millstone Unit 1 PRV during an outage scheduled to begin in April 1989. This examination was performed with an inspection tool placed inside of the vessel and the UDRPS data acquisition system. The inspection tool was based on tooling used routinely at the Swedish and Finnish BWRs. The object of RP C105-1 was to use the information gathered prior to and during the Millstone Unit 1 examination and to perform an inspection accessibility study and prepare a report covering the lessons learned during the examination that would benefit other BWR utilities considering similar inspections. The final report consists of two volumes. This document, Volume 1, describes a 3D model of the Millstone Unit 1 PRV that was based on information obtained from plant drawings. The model is a useful tool for visualizing areas of limited access from both inside and outside of the RPV. 5 refs., 38 figs., 4 tabs

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

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.

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.)

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)

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)

Validated automated ultrasonic inspections of the Sizewell 'B' reactor pressure vessel

International Nuclear Information System (INIS)

Dikstra, B.J.; Farley, J.M.

1992-01-01

Automated ultrasonic inspection was applied extensively during manufacture of the RPV for Sizewell 'B'. This was an important element of the safety case presented at the Sizewell 'B' public enquiry. This requirement reflected concern in the United Kingdom as to the effectiveness and reliability of ultrasonic inspections. By applying automated inspections in addition to the manual ultrasonic inspection carried out by the vessel manufacturer, the overall reliability of the inspection of the vessel would be considerably enhanced. The automated inspections carried out in the manufacturer's workshops were termed 'automated shop inspections' (ASIs). The ASIs were carried out in two contracts: the first to inspect the component forgings of the RPV, the second to inspect the pressure retaining welds. (author)

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

Heavy-section steel irradiation program summary

International Nuclear Information System (INIS)

Corwin, W.R.; Nanstad, R.K.; Iskander, S.K.; Haggag, F.M.

1992-01-01

Since a failure of the RPV carries the potential of major contamination release and severe accident, it is imperative to safe reactor operation to understand and be able to accurately predict failure models of the vessel material. 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 RPVs. The program includes the direct continuation of irradiation studies previously conducted within the Heavy-Section Steel Technology Program augmented by enhanced examinations 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 including fracture toughness (K Ic and J Ic ), crack-arrest toughness (K Ia ), ductile tearing resistance (dJ/da), Charpy V-notch impact energy, dropweight nil-ductility temperature (NDT), and tensile properties. Models based on observations of radiation-induced microstructural changes using field ion and high-resolution transmission electron microscopy provide a firmer basis for extrapolating the measured changes in fracture properties to wider ranges of irradiation conditions. The principal materials examined within the HSSI Program are highcopper welds since their postirradiation properties are most frequently limiting in the continued safe operation of commercial RPVs. In addition, a limited effort will focus on stainless steel weld overlay cladding, typical of that used on the inner surface of RPVs, since its postirradiation fracture properties have the potential for strongly affecting the extension of small surface flaws during overcooling transients. (orig./GL)

Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

International Nuclear Information System (INIS)

Boros, I.; Aszodi, A.

2003-01-01

The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

Energy Technology Data Exchange (ETDEWEB)

Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

2017-04-01

Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

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

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

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)

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)

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

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

Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

Science.gov (United States)

Krasikov, E.

2015-04-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of 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. There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment. The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, 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.

Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

International Nuclear Information System (INIS)

Krasikov, E

2015-01-01

As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of 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.There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment.The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, 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. (paper)

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.

Determination of neutron fluence and radiation brittleness temperature of WWER-440 and WWER-1000 pressure vessels in Kozloduy NPP

International Nuclear Information System (INIS)

Ilieva, K.; Apostolov, T.; Belousov, S.; Petrova, T.; Antonov, S.; Ivanov, K.; Prodanova, R.

1993-01-01

In Units 1-4 of Kozloduy NPP (WWER-440/230), the weld 4 of RPV undergoes the most severe irradiation embrittlement. Neither witness-samples, nor detectors are designed for these reactors. Transport calculations of fast neutron fluence on WWER-440 RPV and ex-vessel measurements by threshold activation detectors are the primary means for adequate assessment of metal state and for prognosis concerning the reactor life span. In WWER-1000 reactors (Units 5-6) the maximum neutron fluence occurs on the weld 3. The systematical observation of metal state is performed through witness-samples and threshold activation detectors ( 54 Fe (n,p), 63 Cu (n,α), 93 Nb (n,n')) placed above the reactor top edge and at the first vessel ring level. There are big differences in energy spectrum and integral neutron flux falling onto the weld 3, the RPV base metal and the staff detectors. This requires additional neutron measurements in the air gap between the RPV and the thermal insulation. (author)

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)

Benchmarking of the computer code and the thirty foot side drop analysis for the Shippingport (RPV/NST package)

International Nuclear Information System (INIS)

Bumpus, S.E.; Gerhard, M.A.; Hovingh, J.; Trummer, D.J.; Witte, M.C.

1989-01-01

This paper presents the benchmarking of a finite element computer code and the subsequent results from the code simulating the 30 foot side drop impact of the RPV/NST transport package from the decommissioned Shippingport Nuclear Power Station. The activated reactor pressure vessel (RPV), thermal shield, and other reactor external components were encased in concrete contained by the neutron shield tank (NST) and a lifting skirt. The Shippingport RPV/NST package, a Type B Category II package, weighs approximately 900 tons and has 17.5 ft diameter and 40.7 ft. length. For transport of the activated components from Shippingport to the burial site, the Safety Analysis Report for Packaging (SARP) demonstrated that the package can withstand the hypothetical accidents of DOE Order 5480.3 including 10 CFR 71. Mathematical simulations of these accidents can substitute for actual tests if the simulated results satisfy the acceptance criteria. Any such mathematical simulation, including the modeling of the materials, must be benchmarked to experiments that duplicate the loading conditions of the tests. Additional confidence in the simulations is justified if the test specimens are configured similar to the package

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.)

Quantification of the toughness distribution in a heavy section submerged arc multilayer reactor pressure vessel weldment

International Nuclear Information System (INIS)

Cerjak, H.; Prader, R.

1999-01-01

In a working procedure qualification test weld representing a heavy section circumferential reactor pressure vessel (RPV) weld tested in 1968, lower toughness values were observed in the top layer region compared to those found in the filler region. Gleeble simulation, extensive microscopic evaluation, diligent Charpy V-notch testing and modelling of the bead sequence and distribution of alloying elements was applied to explain this effect. It could be revealed that the microstructure of the weld metal is the most important factor influencing the toughness. When an 'as welded' microstructure is partly or fully reaustenitised by the adjacent multilayer beads, the microstructure transforms and the toughness increases. In the filler region, 85% of the cross-section consists from transformed microstructure, whereas in the top layer only 20% are transformed. It is quite evident that, accidentally, the notch tip of Charpy samples in 1968 were placed in untransformed microstructures. When the top layer on the inner surface of the RPV is weld cladded by austenitic stainless steel, full transformation occurs and the toughness representing the filler region can be taken into account for safety evaluations. (orig.)

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)

Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

International Nuclear Information System (INIS)

Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I.

1999-12-01

One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (author)

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

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

United States Department of Energy projects related to reactor pressure vessel annealing optimization

International Nuclear Information System (INIS)

Rosinski, S.T.; Nakos, J.T.

1993-01-01

Light water reactor pressure vessel (RPV) material properties reduced by long-term exposure to neutron irradiation can be recovered through a thermal annealing treatment. This technique to extend RPV life, discussed in this report, provides a complementary approach to analytical methodologies to evaluate RPV integrity. RPV annealing has been successfully demonstrated in the former Soviet Union and on a limited basis by the US (military applications only). The process of demonstrating the technical feasibility of annealing commercial US RPVs is being pursued through a cooperative effort between the nuclear industry and the US Department of Energy (USDOE) Plant Lifetime Improvement (PLIM) Program. Presently, two projects are under way through the USDOE PLIM Program to demonstrate the technical feasibility of annealing commercial US RPVS, (1) annealing re-embrittlement data base development and (2) heat transfer boundary condition experiments

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.)

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)

Apparative developments for inservice inspections of reactor pressure vessels

International Nuclear Information System (INIS)

Bohn, H.; Ruthrof, K.; Barbian, O.A.; Kappes, W.; Neumann, R.; Stanger, H.K.

1987-01-01

Emphasizing PWR pressure vessel (RPV) inspections, recent developments of new generations of automated and mechanized ultrasonic inspection equipment are presented. Starting from general equipment design and inservice implenentation criteria, specific examples are given. Main attention is directed to equipment realization of phased array and ALOK inspection techniques, especially in their combination. Refined aspects of subsequent computer processing and evaluation of defect detection data are described. Analytical features and potential for further developments become evident. Remote controlled RPV inspections are stressed by describing a new generation of central mast manipulators, forming an integral part of total inservice inspection system. (orig./HP)

Master curve characterization of the fracture toughness behavior in SA508 Gr.4N low alloy steels

Energy Technology Data Exchange (ETDEWEB)

Lee, Ki-Hyoung, E-mail: shirimp@kaist.ac.k [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of); Kim, Min-Chul; Lee, Bong-Sang [Nuclear Materials Research Division, KAERI, Daejeon 305-353 (Korea, Republic of); Wee, Dang-Moon [Department of Materials Science and Engineering, KAIST, Daejeon 305-701 (Korea, Republic of)

2010-08-15

The fracture toughness properties of the tempered martensitic SA508 Gr.4N Ni-Mo-Cr low alloy steel for reactor pressure vessels were investigated by using the master curve concept. These results were compared to those of the bainitic SA508 Gr.3 Mn-Mo-Ni low alloy steel, which is a commercial RPV material. The fracture toughness tests were conducted by 3-point bending with pre-cracked charpy (PCVN) specimens according to the ASTM E1921-09c standard method. The temperature dependency of the fracture toughness was steeper than those predicted by the standard master curve, while the bainitic SA508 Gr.3 steel fitted well with the standard prediction. In order to properly evaluate the fracture toughness of the Gr.4N steels, the exponential coefficient of the master curve equation was changed and the modified curve was applied to the fracture toughness test results of model alloys that have various chemical compositions. It was found that the modified curve provided a better description for the overall fracture toughness behavior and adequate T{sub 0} determination for the tempered martensitic SA508 Gr.4N steels.

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

Fracture assessment of weld material from a full-thickness clad RPV shell segment

International Nuclear Information System (INIS)

Keeney, J.A.; Bass, B.R.; McAfee, W.J.

1996-01-01

Fracture analysis was applied to full-thickness clad beam specimens containing shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPV) at beginning of life. The beam specimens were fabricated from a section of an RPV wall (removed from a canceled nuclear plant) that includes weld, plate, and clad material. Metallurgical factors potentially influencing fracture toughness for shallow cracks in the beam specimens include gradients of material properties and residual stresses due to welding and cladding applications. Fracture toughness estimates were obtained from load vs load-line displacement and load vs crack-mouth-opening displacement data using finite-element methods and estimation schemes based on the η-factor method. One of the beams experienced a significant amount of precleavage stable ductile tearing. Effects of precleavage tearing on estimates of fracture toughness were investigated using continuum damage models. Fracture toughness results from the clad beam specimens were compared with other deep- and shallow-crack single-edge notch bend (SENB) data generated previously from A533 Grade B plate material. Range of scatter for the clad beam data is consistent with that from the laboratory-scale SENB specimens tested at the same temperature

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

Development of Deterministic and Probabilistic Fracture Mechanics Analysis Code PROFAS-RV for Reactor Pressure Vessel - Progress of the Work

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Min; Lee, Bong Sang [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, a deterministic/probabilistic fracture mechanics analysis program for reactor pressure vessel, PROFAS-RV, is developed. This program can evaluate failure probability of RPV using recent radiation embrittlement model of 10CFR50.61a and stress intensity factor calculation method of RCC-MRx code as well as the required basic functions of PFM program. Applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors, and others which can improve fracture mechanics assessment of RPV are introduced. The purpose of this study is to develop a probabilistic fracture mechanics (PFM) analysis program for RPV considering above modification and application of newly developed models and calculation methods. In this paper, it deals with the development progress of the PFM analysis program for RPV, PROFAS-RV. The PROFAS-RV is being tested with other codes, and it is expected to revise and upgrade by reflecting the latest model and calculation method continuously. These efforts can minimize the uncertainty of the integrity evaluation for the reactor pressure vessel.

Reactor pressure vessel structural integrity research

International Nuclear Information System (INIS)

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

1994-01-01

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

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)

Analysis of the effect of implemented low temperature overpressure regimes on the reactor pressure vessel resistance to damage

International Nuclear Information System (INIS)

Pistora, V.

1995-12-01

The temperature and stress fields of the Dukovany WWER-440 reactor pressure vessel (RPV) were calculated based on a two-dimensional model using the finite element method. Two pressurized thermal shock events occurred at Dukovany in 1992: the temperature in 3 loops dropped rapidly while the primary circuit was fully pressurized. The calculation revealed that the first event was intolerable with respect to the RPV resistance to brittle fracture; had the two events occurred towards the end of the RPV lifetime, both would have been intolerable. (M.D.). 6 tabs., 15 figs., 6 refs

Reactor pressure vessel stud management automation strategies

International Nuclear Information System (INIS)

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

1992-01-01

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

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

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 ...

Formation of nano sized ODS clusters in mechanically alloyed NiAl-(Y,Ti,O) alloys

International Nuclear Information System (INIS)

Kim, Yong Deog; Bae Seong Man; Wirth, Brian D.

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

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

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

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)

Improvement of remote control system of automatic ultrasonic equipment for inspection of reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Cheong, Yong Moo; Jung, H. K.; Joo, Y. S.; Koo, K. M.; Hyung, H.; Sim, C. M.; Gong, U. S.; Kim, S. H.; Lee, J. P.; Rhoo, H. C.; Kim, M. S.; Ryoo, S. K.; Choi, C. H.; Oh, K. I

1999-12-01

One of the important issues related to the nuclear safety is in-service inspection of reactor pressure vessel (RPV). A remote controlled automatic ultrasonic method is applied to the inspection. At present the automatic ultrasonic inspection system owned by KAERI is interrupted due to degradation of parts. In order to resume field inspection new remote control system for the equipment was designed and installed to the existing equipment. New ultrasonic sensors and their modules for RPV inspection were designed and fabricated in accordance with the new requirements of the inspection codes. Ultrasonic sensors were verified for the use in the RPV inspection. (autho0008.

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.

Preparation of the Shippingport reactor pressure vessel shipping package

International Nuclear Information System (INIS)

Yannitell, D.M.

1988-01-01

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

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)

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

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

Fracture mechanics assessment of surface and sub-surface cracks in the RPV under non-symmetric PTS loading

Energy Technology Data Exchange (ETDEWEB)

Keim, E; Shoepper, A; Fricke, S [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

1997-09-01

One of the most severe loading conditions of a reactor pressure vessel (rpv) under operation is the loss of coolant accident (LOCA) condition. Cold water is injected through nozzles in the downcomer of the rpv, while the internal pressure may remain at a high level. Complex thermal hydraulic situations occur and the fluid and downcomer temperatures as well as the fluid to wall heat transfer coefficient at the inner surface are highly non-linear. Due to this non-symmetric conditions, the problem is investigated by three-dimensional non-linear finite element analyses, which allow for an accurate assessment of the postulated flaws. Transient heat transfer analyses are carried out to analyze the effect of non-symmetrical cooling of the inner surface of the pressure vessel. In a following uncoupled stress analysis the thermal shock effects for different types of defects, surface flaws and sub-surface flaws are investigated for linear elastic and elastic-plastic material behaviour. The obtained fracture parameters are calculated along the crack fronts. By a fast fracture analysis the fracture parameters at different positions along the crack front are compared to the material resistance. Safety margins are pointed out in an assessment diagram of the fracture parameters and the fracture resistance versus the transient temperature at the crack tip position. (author). 4 refs, 10 figs.

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

Load bearing capacities and elastic-plastic behavior of reactor vessel internals

International Nuclear Information System (INIS)

Watanabe, Keita; Nagase, Ryuichi

2017-01-01

Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

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

PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

Science.gov (United States)

Leclercq, Sylvain; Lidbury, David; Van Dyck, Steven; Moinereau, Dominique; Alamo, Ana; Mazouzi, Abdou Al

2010-11-01

In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young

PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

Energy Technology Data Exchange (ETDEWEB)

Leclercq, Sylvain, E-mail: sylvain.leclercq@edf.f [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France); Lidbury, David [SERCO Assurance - Walton House, 404 Faraday Street, Birchwood Park, Warrington, Cheshire WA3 6GA (United Kingdom); Van Dyck, Steven [SCK-CEN, Nuclear Material Science, Boeretang 200, BE, 2400 Mol (Belgium); Moinereau, Dominique [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France); Alamo, Ana [CEA Saclay, DEN/DSOE, 91191 Gif-sur-Yvette (France); Mazouzi, Abdou Al [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France)

2010-11-01

In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations... from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach

Intergranular fracture stress and phosphorus grain boundary segregation of a Mn-Ni-Mo steel

International Nuclear Information System (INIS)

Naudin, C.; Frund, J.M.; Pineau, A.

1999-01-01

Nuclear Reactor Pressure Vessel (RPV) steel A508 class 3 which is a low alloyed steel is not usually sensitive to reversible temper embrittlement when properly heat treated. However heterogeneous zones may be present in particular near the inner side of the vessel. These zones result from the segregation of the alloying elements (C, Mn, Ni, Mo) and impurities (S, P) taking place during solidification of the material. They are called segregated zones (or ghost lines). They can reach 2 mm thick along the radius and 30 mm long through the circumferential direction. Their susceptibility to reversible temper embrittlement is mainly due to grain boundary phosphorus segregation triggering brittle intergranular fracture when the material is tested at low temperature. In this material like in other steels the influence of some other alloying elements (Mo, Mn...) is clearly significant and should also be taken into account. But phosphorus effect has proved to be predominant. The aim of the present study is therefore to find out a quantitative relationship between grain boundary phosphorus segregation and critical intergranular fracture stress. A synthetic steel with a chemical composition representative of an average segregated zone was prepared for the present study. A number of heat treatments were applied to reach different embrittlement conditions. Then brittle fracture properties were obtained by performing cryogenic fracture tests on notched tensile specimens while the corresponding grain boundary phosphorus levels were measured by Auger electron spectroscopy. Systematic fractographic observations were carried out. Moreover an attempt to determine the influence of temperature on the critical intergranular fracture stress was made

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)

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)

Application of the RTNDT- and RTT0- concept for the Borssele RPV considering 60 years of operation

Energy Technology Data Exchange (ETDEWEB)

Barthelmes, J.; Keim, E.; Hein, H. [AREVA NP Gmbh (Germany); Jong, A. de [EPZ Kerncentrale Borssele (Netherlands)

2011-07-01

The nuclear power plant (NPP) Borssele started operation in 1973 and was designed for operation until 2014. In order to operate the plant beyond 2013 an assessment for long term operation (LTO) for 60 years was performed. For experimental validation of the RPV irradiation behavior, two irradiation surveillance programs, each consisting of one unirradiated and two irradiated sets, were implemented. Each set consists of capsules with representative material test specimens from the RPV core belt-line region, the base metal (BM) rings 03 and 04 and the weld metal (WM) W 03. The first surveillance program is already evaluated and was designed to cover safe operation for 40 years. With the test results from the two irradiation sets of the first surveillance program and from irradiation data of similar RPV steels a prediction of the adjusted reference temperatures at end of life (EoL), covering 60 years of operation, was carried out. The corresponding maximum accumulated theoretical fast neutron fluence (E> 1 MeV) at the inner RPV wall was calculated to 3.22 E+19 n/cm{sup 2} and 3.40 E+19 n/cm{sup 2}, considering no mixed oxide fuel (MOX) and anticipated MOX core loading management, respectively. The leading material in terms of irradiation induced aging is the WM with an adjusted reference temperature of 18 C according to the RTNDT concept and of 3 C according to the Master curve concept, respectively. The results have large safety margins to the KTA limit curve according to the German safety standard KTA 3203. The predicted data will be subject of experimental confirmation in a few years by the test data of the two irradiation sets of the second surveillance program. (authors)

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-07-01

A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

Assessment of the fracture behavior of weld material from a full-thickness clad RPV shell segment

International Nuclear Information System (INIS)

Bass, B.R.; Keeney, J.A.; McAfee, W.J.

1995-01-01

A testing program is described that utilizes full-thickness clad beam specimens to quantify fracture toughness for shallow cracks in material for which metallurgical conditions are prototypic of those found in reactor pressure vessels (RPVs). The beam specimens are fabricated from a section of an RPV shell (removed from a canceled nuclear plant) that includes weld, plate, and clad material. A summary of the testing program includes a description of the specimen geometry, material properties, the testing procedure, and the experimental results from three specimens. The yield strength of the weld material was determined to be 36% higher than the base material. The high yield strength for prototypic weld material may be implications for RPV integrity assessments. Fracture toughness data from three clad beam specimens are compared with other shallow- and deep-crack beam cruciform data generated previously from A 533 Grade B plate material. Difficulties with interpreting lower-bound fracture toughness curves constructed from the shallow-crack data are essentially resolved by adopting a single normalizing temperature parameter, namely, the nil-ductility transition temperature (NDT)

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.

Summary of flow and heat transfer in RPV under PTS

International Nuclear Information System (INIS)

Lu Donghua; Wang Haijun; Chen Tingkuan; Luo Yushan

2003-01-01

PTS under loss of coolant accident (LOCA) has great effect on the safety of RPV. Many research works focusing on flow and heat transfer in RPV under PTS have been done in developed countries for many years, and a lot of results have been got both on experiment and numerical simulation. The safety of nuclear power plant is enhanced greatly by these research works. With the developing of nuclear power technology in China, RPV integration under PTS has been studied. The author summarizes research works at home and abroad in recent years. The problems existed in present work and research direction in the future are discussed

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.

A physically-based constitutive model for SA508-III steel: Modeling and experimental verification

International Nuclear Information System (INIS)

Dong, Dingqian; Chen, Fei; Cui, Zhenshan

2015-01-01

Due to its good toughness and high weldability, SA508-III steel has been widely used in the components manufacturing of reactor pressure vessels (RPV) and steam generators (SG). In this study, the hot deformation behaviors of SA508-III steel are investigated by isothermal hot compression tests with forming temperature of (950–1250)°C and strain rate of (0.001–0.1)s −1 , and the corresponding flow stress curves are obtained. According to the experimental results, quantitative analysis of work hardening and dynamic softening behaviors is presented. The critical stress and critical strain for initiation of dynamic recrystallization are calculated by setting the second derivative of the third order polynomial. Based on the classical stress–dislocation relation and the kinetics of dynamic recrystallization, a two-stage constitutive model is developed to predict the flow stress of SA508-III steel. Comparisons between the predicted and measured flow stress indicate that the established physically-based constitutive model can accurately characterize the hot deformations for the steel. Furthermore, a successful numerical simulation of the industrial upsetting process is carried out by implementing the developed constitutive model into a commercial software, which evidences that the physically-based constitutive model is practical and promising to promote industrial forging process for nuclear components

A physically-based constitutive model for SA508-III steel: Modeling and experimental verification

Energy Technology Data Exchange (ETDEWEB)

Dong, Dingqian [National Die & Mold CAD Engineering Research Center, Shanghai Jiao Tong University, 1954 Huashan Rd., Shanghai 200030 (China); Chen, Fei, E-mail: feechn@gmail.com [National Die & Mold CAD Engineering Research Center, Shanghai Jiao Tong University, 1954 Huashan Rd., Shanghai 200030 (China); Department of Mechanical, Materials and Manufacturing Engineering, University of Nottingham, Nottingham NG7 2RD (United Kingdom); Cui, Zhenshan, E-mail: cuizs@sjtu.edu.cn [National Die & Mold CAD Engineering Research Center, Shanghai Jiao Tong University, 1954 Huashan Rd., Shanghai 200030 (China)

2015-05-14

Due to its good toughness and high weldability, SA508-III steel has been widely used in the components manufacturing of reactor pressure vessels (RPV) and steam generators (SG). In this study, the hot deformation behaviors of SA508-III steel are investigated by isothermal hot compression tests with forming temperature of (950–1250)°C and strain rate of (0.001–0.1)s{sup −1}, and the corresponding flow stress curves are obtained. According to the experimental results, quantitative analysis of work hardening and dynamic softening behaviors is presented. The critical stress and critical strain for initiation of dynamic recrystallization are calculated by setting the second derivative of the third order polynomial. Based on the classical stress–dislocation relation and the kinetics of dynamic recrystallization, a two-stage constitutive model is developed to predict the flow stress of SA508-III steel. Comparisons between the predicted and measured flow stress indicate that the established physically-based constitutive model can accurately characterize the hot deformations for the steel. Furthermore, a successful numerical simulation of the industrial upsetting process is carried out by implementing the developed constitutive model into a commercial software, which evidences that the physically-based constitutive model is practical and promising to promote industrial forging process for nuclear components.

Development and deployment of BARC Vessel Inspection System (BARVIS) for TAPS-1 and 2

International Nuclear Information System (INIS)

Singh, Jit Pal; Ranjon, R.; Kulkarni, M.P.; Soni, N.L.; Patel, R.J.

2016-01-01

As per regulatory requirements, inspection of welds in Reactor Pressure Vessel (RPV) is necessary for further continuing operation of TAPS - 1 and 2. Upper shell longitudinal welds of RPV which were seen as inaccessible up till now have been inspected first time since operation of the reactors by deploying Weld Inspection Manipulator (WIM) in Unit-1 in August 2012. Subsequently Unit-2 and again Unit-1 upper shell welds were inspected with upgraded versions of WIM in Feb 2013 and March 2015 respectively. Inspection of upper shell welds paves the way for more challenging inspection of beltline region welds. These welds are accessible only from Inside Diameter (JD) surface through a narrow annular gap of 25 mm between RPV wall and thermal shield by managing obstructions due to core internals. BARC Vessel Inspection System (BARVIS) for inspection of beltline region welds from inner side of the RPV was designed, manufactured, tested and qualified for sending scanning probes in to the annular gap of 25mm as per RPV engineering drawings and also based on actual gap measured in Unit-2. Annular gap measurement was done in Units-1 before deployment of BARVIS during 25 th Refuelling Outage (RFO) in Unit-1. Contrary to the expectation, the annular gap was found less and inspection of beltline region with BARVIS in Unit-1 could not be done. Finally during RFO in January 2016 of Unit-2, BARVIS has been successfully deployed for beltline region welds inspection. BARVIS mainly consists of manipulator, its operating system and data acquisition system. Data acquisition system and data analysis are not covered in this report. (author)

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

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

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.)

Reactor pressure vessel life cycle management at the Calvert Cliffs Nuclear Power Plant

International Nuclear Information System (INIS)

Doroshuk, B.W.; Bowman, M.E.; Henry, S.A.; Pavinich, W.A.; Lapides, M.E.

1993-01-01

Life Cycle Management (LCM) seeks to manage the aging process of important systems, structures, and components during licensed operation. The goal of Baltimore Gas and Electric Company's (BG and E) Life Cycle Management Program is to assure attainment of 40 years of operation and to preserve the option of an additional 20 years of operation for the Calvert Cliffs Nuclear Power Plant (CCNPP). Since the reactor pressure vessel (RPV) has been identified as one of the most critical components with regard to long-term operation of a nuclear power plant, BG and E initiated actions to manage life limiting or aging issues for the CCNPP RPVs. To achieve long-term operation, technical RPV issues must be effectively managed. This paper describes methods BG and E uses for managing RPV age-related degradation. (author)

Pressurized thermal shock. Thermo-hydraulic conditions in the CNA-I reactor pressure vessel

International Nuclear Information System (INIS)

Ventura, Mirta A.; Rosso, Ricardo D.

2002-01-01

In this paper we analyze several reports issued by the Utility (Nucleo Electrica S.A.) and related to Reactor Pressure Vessel (RPV) phenomena in the CNA-I Nuclear Power Plant. These analyses are aimed at obtaining conclusions and establishing criteria ensuring the RPV integrity. Special attention was given to the effects ECCS cold-water injection at the RPV down-comer leading to pressurized thermal shock scenarios. The results deal with hypothetical primary system pipe breaks of different sizes, the inadvertent opening of the pressurizer safety valve, the double guillotine break of a live steam line in the containment and the inadvertent actuation pressurizer heaters. Modeling conditions were setup to represent experiments performed at the UPTF, under the hypothesis that they are representative of those that, hypothetically, may occur at the CNA-I. No system scaling analysis was performed, so this assertion and the inferred conclusions are no fully justified, at least in principle. The above mentioned studies, indicate that the RPV internal wall surface temperature will be nearly 40 degree. It was concluded that they allowed a better approximation of PTS phenomena in the RPV of the CNA-I. Special emphasis was made on the influence of the ECCS systems on the attained RPV wall temperature, particularly the low-pressure TJ water injection system. Some conservative hypothesis made, are discussed in this report. (author)

Adequacy of Current Equivalent Margins Analysis (EMA) Guidance, Data and Methodologies for 60+ Years of Operation

International Nuclear Information System (INIS)

Server, W.; Hardin, T.; Cipolla, R.; Hall, B.

2015-01-01

In order to assure the structural integrity of reactor pressure vessels (RPVs), the fracture toughness of the ferritic steels used to fabricate the RPV must be shown to be adequate during their entire operating life, including extended license life. The Charpy V-notch (CVN) impact test has been used in the nuclear industry since it uses a small test specimen that can be irradiated in surveillance programs and provides an indirect way of assessing the fracture toughness of RPV steels. The effects of embrittlement typically are characterized by changes to the average Charpy curves measured before and after irradiation: shift of the 30 ft-lb (41 J) index temperature, and decrease in the CVN upper shelf energy (USE). Requirements in the USA for the USE of RPV belt-line materials are codified in Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Appendix G. Before irradiation, USE in the transverse (T-L) orientation for base materials and crack extension in the welding direction for weld materials must be greater than or equal to 75 ft-lb (102 J), and it is not to become less than 50 ft-lb (68 J) due to radiation embrittlement throughout the license of the RPV. If the projected USE of any RPV belt-line steel falls below 50 ft-lb (68 J), the projected value must be demonstrated to provide a margin of safety against ductile fracture equivalent to that required by Appendix G of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI. The analytical evaluation method used is called an equivalent margin analysis (EMA). This paper reviews the current status of EMAs and recommends improvements and clarifications that can be made to meet the needs of extended license life to 80 years. Focus is placed on analytical methodology, material property needs and proper implementation. (authors)

Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

Energy Technology Data Exchange (ETDEWEB)

Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

2017-12-15

This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

Decision process involved in preparing the Shippingport reactor pressure vessel for transport

International Nuclear Information System (INIS)

Murphie, W.E.

1989-01-01

The most significant part of the Shippingport Station Decommissioning Project was the one-piece removal and shipment of the reactor pressure vessel (RPV). Implicit in the RPV transport was the task of qualifying the RPV as a waste package acceptable for shipment. Soon after physical decommissioning began on September 1985, questions regarding the packaging certification and transport of the RPV from Shippingport, Pennsylvania to the US Department of Energy (DOE) Hanford Waste Burial Site necessitated reexamination of several planning assumptions. A complete reassessment of the regulatory requirements governing the RPV shipment resulted in a programmatic decision to obtain a type B(U) Certificate of Compliance and abandon the originally planned US Department of Transportation (DOT) low specific activity (LSA) shipment. The decision process resulting in this conclusion was extensive and involved many organizations and agencies. Incidental to this process, several subtle certification issues were identified that required resolution. Some of these issues involved the definition of LSA material for large packages; interpretation and compliance with DOE, DOT and US Nuclear Regulatory Commission (NRC) regulations for the transport of radioactive material; incorporation of the International Atomic Energy Agency (IAEA) regulations by the Panama Canal; and DOE policy requiring advance notification to states of radioactive waste shipments. 2 figs

Decision process involved in preparing the Shippingport reactor pressure vessel for transport

International Nuclear Information System (INIS)

Murphie, W.E.

1990-01-01

The most significant part of the Shippingport Station Decommissioning Project was the one-piece removal and shipment of the reactor pressure vessel (RPV). Implicit in the RPV transport was the task of qualifying the RPV as a waste package acceptable for shipment. Soon after physical decommissioning began on September, 1985, questions regarding the packaging certification and transport of the RPV from Shippingport, Pennsylvania to the U.S. Department of Energy (DOE) Hanford waste burial site necessitated reexamination of several planning assumptions. A complete reassessment of the regulatory requirements governing the RPV shipment resulting in a programmatic decision to obtain a Type B(U) Certification of Compliance and abandon the originally planned U.S. Department of Transportation (DOT) low specific activity (LSA) shipment. The decision process resulting in this conclusion was extensive and involved many organizations and agencies. Incidental to this process, several subtle certification issues were identified that required resolution. Some of these issues involved the definition of LSA material for large packages; interpretation and compliance with DOE, DOT and U.S. Nuclear Regulatory Commission (NRC) regulations for the transport of radioactive material; incorporation of the International Atomic Energy Agency (IAEA) regulations by the Panama Canal; and DOE policy requiring advance notification to states of radioactive waste shipments

Study on reactor vessel replacement (RVR) for 1100 MW class BWR plants in Japan

International Nuclear Information System (INIS)

Mizutani, J.; Kawamura, S.; Aoki, M.; Mori, T.

2001-01-01

Plant Life Management (PLM) is being studied in Japan, and reactor vessel replacement (RVR) is being considered as one option. Since reactor internals, except for reusable parts, and the reactor pressure vessel (RPV) are replaced, the RVR provides an effective technology for extending the service life of nuclear power plants substantially. At ICONE 7, we reported on the technical viability of the RVR for BWR4-type 800 MWe class plants. This time, we rationalized the RVR method through a study for BWR5-type 1100 MWe class plants to reduce the RVR duration and evaluated the technical viability and the economic efficiency of the method. In addition, we discuss how to dispose of the RPV to complete a scenario of the process from the RVR to its final disposal. (author)

Simultaneous B and L violation: new signatures from RPV-SUSY

International Nuclear Information System (INIS)

Faroughy, Cyrus; Prabhu, Siddharth; Zheng, Bob

2015-01-01

Studies of R-parity violating (RPV) supersymmetry typically assume that nucleon stability is protected by approximate baryon number (B) or lepton number (L) conservation. We present a new class of RPV models that violate B and L simultaneously (BLRPV), without inducing rapid nucleon decay. These models feature an approximate Z 2 e ×Z 2 μ ×Z 2 τ flavor symmetry, which forbids 2-body nucleon decay and ensures that flavor antisymmetric LLE c couplings are the only non-negligible L-violating operators. Nucleons are predicted to decay through N→Keμν and n→eμν; the resulting bounds on RPV couplings are rather mild. Novel collider phenomenology arises because the superpartners can decay through both L-violating and B-violating couplings. This can lead to, for example, final states with high jet multiplicity and multiple leptons of different flavor, or a spectrum in which depending on the superpartner, either B or L violating decays dominate. BLRPV can also provide a natural setting for displaced ν̃→μe decays, which evade many existing collider searches for RPV supersymmetry.

Ultrasonic phased array examination of circumferential weld joint in reactor pressure vessel of BWR

Energy Technology Data Exchange (ETDEWEB)

Nanekar, Paritosh, E-mail: pnanekar@barc.gov.in [Quality Assurance Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jothilakshmi, N. [Quality Assurance Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Jayakumar, T. [Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

2013-12-15

Highlights: • Phased array technique developed for weld joint inspection in BWR pressure vessel. • Simulation studies were carried out for conventional and phased array probe. • Conventional ultrasonic test shows in-adequate weld coverage and poor resolution. • Focused sound beam in phased array results in good resolution and sensitivity. • Ultrasonic phased array technique is validated on mock-up with reference defects. - Abstract: The weld joints in the reactor pressure vessel (RPV) of Boiling Water Reactors (BWR) are required to be examined periodically for assurance of structural integrity. Ultrasonic phased array examination technique has been developed in authors’ laboratory for inspection of the top flange to shell circumferential weld joint in RPV of BWRs, which are in operation in India since the late 1960s. The development involved detailed simulation studies for computation of focal laws followed by validation on mock-up. The paper brings out the limitations of the conventional ultrasonic technique and how this can be overcome by the phased array approach for the weld joint under consideration. The phased array technique was successfully employed for field examination of this weld joint in RPV during the re-fuelling outage.

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.

In-vessel core debris retention through external flooding of the reactor pressure vessel. SCDAP/RELAP5 assessment for the SBWR lower head

International Nuclear Information System (INIS)

Heel, A.M.J.M. van.

1995-09-01

In this report the results are discussed from various analyses on the feasibility and phenomenology of the External Flooding (EF) concept for an SBWR lower head, filled with a large heat generating corium mass. In applying External Flooding as an accident management strategy after or during core melt down, the lower drywell is filled with water up to a level where a large portion of the Reactor Pressure Vessel (RPV) is flooded. The purpose of this method is to establish cooling of the vessel wall, that is challenged by the heat load resulting from the corium, in such a way that its structural integrity if not endangered. The analysis discussed in this report focus on the thermal response of the vessel wall and the ex-vessel boiling processes under the conditions described above. For these analyses the SCDAP/REALP5 MOD 3.1 code was used. The major outcome of the calculations is, that a major part of the vessel wall remains well below themelting temperature of carbon steel, as long as flooding of the external surface of the lower head is established. The SCDAP/RELAP5 analyses indicated that low-quality Critical Heat Flux (CHF) was not exceeded, under all the conditions that had been tested. However, a comaprison of the heat fluxes, as calculated in RELAP5, with the CHF values obtained from the Zuber correlation and the Vishnev correction factor (for boiling at inclined surfaces) proved that CHF values, based on these criteria, were exceeded in several surface points of the lower head mesh. The correlations, as resident in the current version of RELAP5 MOD 3.1, might lead to over-estimation of CHF for the EF analyses discussed in this report. The use of the more conservative Zuber correlation with the Vishnev correction factor is recommended for EF analyses. (orig.)

Identification and evaluation of PWR in-vessel severe accident management strategies

International Nuclear Information System (INIS)

Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

1992-03-01

This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

Atucha I nuclear power plant azimuthal ex-vessel flux profile evaluation

International Nuclear Information System (INIS)

Ferraro, Diego

2008-01-01

Irradiation damage in RPV (Reactor Pressure Vessel) in nuclear power plants is a key parameter to be analyzed in order to assess the plant integrity up to end of life and planning for a possible plant life extension. In this work a neutronic model in MCNP that represents a sector of 30 degrees of the Atucha I power plant nucleus has been consolidated with the results of an ex-vessel dosimetry made in the outer surface of the RPV s power plant in order to analyse the irradiation damage through the dpa rate. A strong dependents of the maximum point of damage with the loading of a peripheral channel was found, so a mitigation strategy was proposed, which is basically to empty this channel and its analogs in the rest of the nucleus. Analysing this second case a notable decrease of the damage is found in the zone considerated on the model (shown through the drop of de dpa rate in the zone). [es

Modelling of in-vessel retention after relocation of corium into the lower plenum - Evaluation of the temperature field and of the viscoplastic deformation of the vessel wall. Reactor safety research, project No.:150 1254 - Final report; Beitrag zur Modellierung der Schmelzerueckhaltung im RDB nach Verlagerung von Corium in das untere Plenum - Berechnung des Temperaturfeldes und der viskoplastischen Verformung der Behaelterwand. Reaktorsicherheitsforschung, Vorhaben-Nr.: 150 1254 - Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Altstadt, E.; Willschuetz, H.G. [Forschungszentrum Rossendorf e.V. (FZR), Dresden (Germany)

2005-01-01

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. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute Of Safety Research of the FZR a finite element model has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal hydraulic and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test series representing the RPV of a PWR in the scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. The results of the calculations can be summarised as follows: The creeping process is caused by the simultaneous presence of high temperature (>600 C) and pressure (>1 MPa). The hot focus region is the most endangered zone exhibiting the highest creep strain rates. The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position. The failure time can be predicted with an uncertainty of 20 to 25%. This uncertainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. The

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)

Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

2016-01-01

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study

Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

2016-10-15

If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.

Simultaneous B and L violation: new signatures from RPV-SUSY

Energy Technology Data Exchange (ETDEWEB)

Faroughy, Cyrus [Department of Physics and Astronomy, Johns Hopkins University,Baltimore, MD 21218 (United States); Prabhu, Siddharth [Department of Physics, Yale University,New Haven, CT 06511 (United States); Zheng, Bob [Michigan Center for Theoretical Physics, University of Michigan,Ann Arbor, MI 48109 (United States)

2015-06-11

Studies of R-parity violating (RPV) supersymmetry typically assume that nucleon stability is protected by approximate baryon number (B) or lepton number (L) conservation. We present a new class of RPV models that violate B and L simultaneously (BLRPV), without inducing rapid nucleon decay. These models feature an approximate Z{sub 2}{sup e}×Z{sub 2}{sup μ}×Z{sub 2}{sup τ} flavor symmetry, which forbids 2-body nucleon decay and ensures that flavor antisymmetric LLE{sup c} couplings are the only non-negligible L-violating operators. Nucleons are predicted to decay through N→Keμν and n→eμν; the resulting bounds on RPV couplings are rather mild. Novel collider phenomenology arises because the superpartners can decay through both L-violating and B-violating couplings. This can lead to, for example, final states with high jet multiplicity and multiple leptons of different flavor, or a spectrum in which depending on the superpartner, either B or L violating decays dominate. BLRPV can also provide a natural setting for displaced ν̃→μe decays, which evade many existing collider searches for RPV supersymmetry.

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.)

The inclusion of weld residual stress in fracture margin assessments of embrittled nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Dickson, T.L.; Bass, B.R.; McAfee, W.J.

1998-01-01

Analyses were performed to determine the impact of weld residual stresses in a reactor pressure vessel (RPV) on (1) the generation of pressure temperature (P-T) curves required for maintaining specified fracture prevention margins during nuclear plant startup and shutdown, and (2) the conditional probability of vessel failure due to pressurized thermal shock (PTS) loading. The through wall residual stress distribution in an axially oriented weld was derived using measurements taken from a shell segment of a canceled RPV and finite element thermal stress analyses. The P-T curve derived from the best estimate load analysis and a t / 8 deep flaw, based on K Ic , was less limiting than the one derived from the current methodology prescribed in the ASME Boiler and Pressure Vessel Code. The inclusion of the weld residual stresses increased the conditional probability of cleavage fracture due to PTS loading by a factor ranging from 2 to 4

DELPHIN - a new system for testing reactor pressure vessels

International Nuclear Information System (INIS)

Dressler, K.

1999-01-01

The author discusses the question of whether a new concept for testing RPVs is necessary. He concentrates his exposition upon the RPV testing system DELPHIN recently developed by ABB, which has been successfully employed in vessel-interior tests since 1998. The new system reduces both the time required and the financial costs for RPV tests. The tests have become more efficient particularly as a result of new developments in the fields of handling machinery and microelectronics. As an example of the improved quality, the author quotes the ultrasonic system ZAQUS: thanks to the high quality of the ultrasonic data, rapid comparison with the results of earlier repreated tests on the RPV is now possible. Since problems of interpretation did not arise, the overall results in an initial application were available only two hours after the last data recording. The author's verdict: DELPHIN has successfully undergone its 'baptism by fire'; still in need of improvement, he states, is the occupany time of the pond, which is not yet as short as targeted. (orig.) [de

A structural evaluation of the Shippingport reactor pressure vessel for transport impact conditions

International Nuclear Information System (INIS)

Witte, M.C.; Chou, C.K.

1989-01-01

The Shippingport Atomic Power Station in Shippingport, Pennsylvania, is being decommissioned and dismantled. This government-leased property will be returned, in a radiologically safe condition, to its owner. All radioactive material is being removed from the Shippingport Station and transported for burial to the DOE Hanford Reservation in Richland, Washington. The reactor pressure vessel (RPV) will be transported by barge to Hanford. This paper describes an evaluation of the structural response of the RPV to the normal and accident impact test conditions as required by the Code of Federal Regulations. 3 refs., 5 figs., 3 tabs

Coupled structure-fluid analysis for a PWR burst protection design

International Nuclear Information System (INIS)

Huber, A.; Hofmann, H.

1977-01-01

The burst protection designed to withstand hypothetical ruptures which might occur in certain components of the primary circuit including RPV (reactor pressure vessel) rupture mainly consists of cylindrical concrete vessels for the RPV and the steam generators and steel tubing for the primary pipes. A hypothetical RPV failure will result in direct excitation of single components and will lead to complex interactions between all components of the protecting structures, the primary loop, reactor core, core support structures and the coolant. The overall investigations to determine the magnitude of deformations and stresses are summaized. Economical aspects with respect to the investigations are treated biefly. The coupled structure-fluid analysis of the core and core support structure due to horizontal and vertical RPV failure will be presented in detail. Assumptions for the RPV failure modes include vertical, horizontal and screw-shaped rupture of the RPV, the detachment of RPV nozzle as well as other types of failure. On the basis of the failure modes, types of credible extremal load conditions were estimated. For vertical RPV failure modes, loads were applied to a global beam-model consisting of burst protection and primary loop structures. Nonlinear coupling between structural parts was taken into account. The nonsymmetric boundary conditions were taken into account by Fourier-expansion in circumferential direction. The mathematical solution is based on the governing equations for pressure wave propagation in fluids and vibrations in solids. Horizontal rupture of the RPV was assumed to occur in the welding connecting spherical bottom and cylinder. Inertia terms of the fluid were incorporated in the equations of the system

Preliminary assessment of the effects of biaxial loading on reactor pressure vessel structural-integrity-assessment technology

International Nuclear Information System (INIS)

Pennell, W.E.; Bass, B.R.; Bryson, J.W.; Dickson, T.L.; McAfee, W.J.; Merkle, J.G.

1996-01-01

Effects of biaxial loading on shallow-flaw fracture toughness were studied to determine potential impact on structural integrity assessment of a reactor pressure vessel (RPV) under pressurized thermal shock (PTS) transient loading and pressure-temperature (PT) loading produced by reactor heatup and cooldown transients. Biaxial shallow-flaw fracture-toughness tests results were also used to determine the parameter controlling fracture in the transition temperature range, and to develop a related dual-parameter fracture-toughness correlation. Shallow-flaw and biaxial loading effects were found to reduce the conditional probability of crack initiation by a factor of nine when the shallow-flaw fracture-toughness K Jc data set, with biaxial-loading effects adjustments, was substituted in place of ASME Code K Ic data set in PTS analyses. Biaxial loading was found to reduce the shallow-flaw fracture toughness of RPV steel such that the lower-bound curve was located between ASME K Ic and K IR curves. This is relevant to future development of P-T curve analysis procedures. Fracture in shallow-flaw biaxial samples tested in the lower transition temperature range was shown to be strain controlled. A strain-based dual-parameter fracture-toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture

Reactor-vessel-sectioning demonstration

International Nuclear Information System (INIS)

Lundgren, R.A.

1981-09-01

A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

Results from the decontamination of and the shielding arrangements in the reactor pressure vessel in Oskarshamn 1-1994