Relevant bibliographies by topics / Nuclear reactor vessel

Academic literature on the topic 'Nuclear reactor vessel'

Author: Grafiati
Published: 10 December 2022
Last updated: 28 January 2023

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Journal articles on the topic "Nuclear reactor vessel"

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Zhou, Linjun, Jie Dai, Yang Li, Xin Dai, Changsheng Xie, Linze Li, and Liansheng Chen. "Research Progress of Steels for Nuclear Reactor Pressure Vessels." Materials 15, no. 24 (December 8, 2022): 8761. http://dx.doi.org/10.3390/ma15248761.

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The nuclear reactor pressure vessel is an important component of a nuclear power plant. It has been used in harsh environments such as high temperature, high pressure, neutron irradiation, thermal aging, corrosion and fatigue for a long time, which puts forward higher standards for the performance requirements for nuclear pressure vessel steel. Based on the characteristics of large size and wall thickness of the nuclear pressure vessel, combined with its performance requirements, this work studies the problems of forging technology, mechanical properties, irradiation damage, corrosion failure, thermal aging behavior and fatigue properties, and summarizes the research progress of nuclear pressure vessel materials. The influencing factors of microstructures evolution and mechanism of mechanical properties change of nuclear pressure vessel steel are analyzed in this work. The mechanical properties before and after irradiation are compared, and the influence mechanisms of irradiation hardening and embrittlement are also summarized. Although the stainless steel will be surfacing on the inner wall of nuclear pressure vessel to prevent corrosion, long-term operation may cause aging or deterioration of stainless steel, resulting in corrosion caused by the contact between the primary circuit water environment and the nuclear pressure vessel steel. Therefore, the corrosion behavior of nuclear pressure vessels materials is also summarized in detail. Meanwhile, the evolution mechanism of the microstructure of nuclear pressure vessel materials under thermal aging conditions is analyzed, and the mechanisms affecting the mechanical properties are also described. In addition, the influence mechanisms of internal and external factors on the fatigue properties, fatigue crack initiation and fatigue crack propagation of nuclear pressure vessel steel are analyzed in detail from different perspectives. Finally, the development direction and further research contents of nuclear pressure vessel materials are prospected in order to improve the service life and ensure safe service in harsh environment.
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Kondylakis, J. S. "Theoretically and under very special applied conditions a nuclear fission reactor may explode as nuclear bomb." HNPS Proceedings 18 (November 23, 2019): 121. http://dx.doi.org/10.12681/hnps.2558.

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This article/presentation describes a theoretical and applied research in nuclear fission reactor systems. It concerns with theoretical approaches and in very special applied cases consideration where a common nuclear fission reactor system may be considered to explode as nuclear bomb. This research gives critical impacts to the design, operation, management and philosophy of nuclear fission reactors systems. It also includes a sensitivity analysis of a particular applied problem concerning the core melting of a nuclear reactor and its deposit to the bottom of reactor vessel. Specifically, in a typical nuclear fission power reactor system of about 1000 MWe, the nuclear core material (corium) in certain cases can be melted and it may deposited in the bottom of nuclear reactor vessel. So, the nuclear criticality conditions are evaluated for a particular example case(s). Assuming an example composition of melted corium of 98 tones of U238 , 1 tone of U235 , 1 tone Pu239 and 25 tones Fe56 (supporting material) in a 5 m diameter of a finite cylindrical nuclear reactor vessel it is found that it may result in nuclear criticality above the unit. This condition corresponds to Supercritical Fast Nuclear Fission Reactor case, which may under certain very special applied conditions to nuclear explode as nuclear bomb.
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Kantsedalov, V. G., V. P. Samoilenko, and A. T. Toporkov. "Remote checking of nuclear-reactor vessel pipes." Soviet Atomic Energy 62, no. 4 (April 1987): 326–29. http://dx.doi.org/10.1007/bf01123375.

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Kramskoi, A. V., Y. G. Lyudmirsky, M. E. Zhidkov, and M. I. Kramskaia. "On extending the life of nuclear reactors." Journal of Physics: Conference Series 2131, no. 2 (December 1, 2021): 022030. http://dx.doi.org/10.1088/1742-6596/2131/2/022030.

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Abstract To extend the service life of nuclear reactors, witness samples from the shells of the core of the reactor vessel are placed in their core. According to the requirements in force in the industry, the reference samples are loaded into the reactor plant unloaded up to the design stresses. This can lead to a biased assessment of the possible extension of the reactor’s life. In connection with the above, in order to assess the mutual influence of operating factors and the stress-strain state of the base metal and welded joints on embrittlement, the reference specimens must be loaded with a force that causes the maximum possible stresses in the specimens during the operation of the reactor. On the basis of domestic and international experience, a test procedure, design and loading scheme for compact witness samples are proposed for modeling and assessing the mutual influence of operating factors and stress-strain state on the object under study (VVER power reactor vessel). For VVER RPVs, the duration of the additional service life should be confirmed by the justification that by the end of the additional service life, the fracture toughness values of the base metal and metal of the welded seams located in the irradiation zone will allow without destruction to withstand all operational and emergency loads, as well as loads at hydraulic tests.
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Zabusov, Oleg O., Boris A. Gurovich, Evgenia A. Kuleshova, Michail A. Saltykov, Svetlana V. Fedotova, and Alexey P. Dementjev. "Intergranular Embrittlement of Nuclear Reactor Pressure Vessel Steels." Key Engineering Materials 592-593 (November 2013): 577–81. http://dx.doi.org/10.4028/www.scientific.net/kem.592-593.577.

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Service life of VVER-type nuclear reactor is limited by decrease in brittle fracture resistance of reactor pressure vessel produced of low-alloy low-carbon steel under effect of irradiation and/or elevated temperatures. In this work fracture surfaces were studied by Auger-electron spectroscopy in order to estimate the contribution of intergranular embrittlement to the degradation of reactor pressure vessel steels under the influence of operating conditions. It was demonstrated that irradiation induced segregation leads to an increase of P content in grain boundaries that promotes intergranular brittle fracture on fracture surfaces. The similar effect but to a lesser degree was shown in the case of long-term temperature exposure. The grain boundary structure was examined and an effect of carbides located on the grain boundaries is supposed due to increased phosphorus segregation on carbide/matrix interface boundaries.
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Dombrovskii, Leonid A., Vladimir N. Mineev, Anatolii S. Vlasov, Leonid I. Zaichik, Yuri A. Zeigarnik, Andrei B. Nedorezov, and Aleksandr S. Sidorov. "In-vessel corium catcher of a nuclear reactor." Nuclear Engineering and Design 237, no. 15-17 (September 2007): 1745–51. http://dx.doi.org/10.1016/j.nucengdes.2007.03.009.

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Rosinski, S. T. "Nuclear reactor pressure vessel-specific flaw distribution development." Theoretical and Applied Fracture Mechanics 19, no. 2 (November 1993): 133–43. http://dx.doi.org/10.1016/0167-8442(93)90015-4.

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Azhagarason, B., N. Mahendran, Tarun Kumar Mitra, and Prabhat Kumar. "Technological Challenges in Manufacturing of over Dimensional Stainless Steel Components of PFBR." Advanced Materials Research 794 (September 2013): 186–93. http://dx.doi.org/10.4028/www.scientific.net/amr.794.186.

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Prototype Fast Breeder Reactor (PFBR) is liquid sodium cooled, pool type nuclear reactor with generating capacity of 1250 MWt / 500MWe. Reactor assembly consists of many large dimensional components made of special grade austenitic stainless steel material. Safety vessel and Main vessel are torispherical dished end vessels with overall height of 12.8 m and 13.4/12.9 m diameter with thickness ranging from 20 to 40 mm. Vessels approx. 111 / 135 MT with running weld length of 500 & 540 m. Inner vessel and thermal baffles are the internals of reactor assembly made of SS 316LN. Forming of dished end petals, weld overlay on the inside surface, circumference matching between the cylindrical shells, cylindrical shell to dished portion was achieved within the tolerances specified. Due to limitations of transportation, these large sized components were manufactured at PFBR site. This paper discusses the experiences gained during the manufacturing of such over dimensional components at PFBR site in meeting the stringent tolerances on various dimensions and NDE requirements.
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Popov, V., V. Mileikovskyi, and O. O. Tryhub. "Expert express assessment of the impact of heat and mass transfer processes on the residual life of the WWER-1000 reactor vessel due to metal embrittlement." Ventilation, Illumination and Heat Gas Supply 41 (April 12, 2022): 39–49. http://dx.doi.org/10.32347/2409-2606.2022.41.39-49.

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The WWER-1000 reactor is operated at 13 of the 15 operating power units of Ukraine's nuclear power plants (NPPs). Ensuring long-term and safe operation of such reactors is the basis for reliable operation of all 13 Ukrainian nuclear power plants units and the guarantor of Ukraine's energy security. The determining and leading factor influencing the safety and proper residual life of the WWER-1000 reactor vessel is the radiation embrittlement of the reactor steel. The consequences of radiation embrittlement of reactor steel are negatively manifested in emergencies with cooling of the core. This process itself – radiation embrittlement – accumulates constantly and gradually. Therefore, it is important to monitor it by periodically performing ongoing rapid assessments of the brittle strength of the WWER-1000 reactor vessel (along with other factors, including cyclic damage, as discussed in a previous publication). Therefore, it is important to use the calculated express methods of periodic assessment of the brittle strength of the WWER-1000 reactor vessel with guaranteed accuracy. The effectiveness of the approach is supported by low cost of resources – engineering staff, fast and relatively simplified use of computers and software. As an example and confirmation of the applicability of the proposed approach, an expert rapid assessment of the fragile strength and residual life of the reactor vessel of Unit № 1 of the South-Ukrainian Nuclear Power Plant was performed. This takes into account the actual, passport characteristics of its metal. The negative impact of the rigid regime with cooling of the WWER-1000 reactor of Unit № 1, not taken into account by the operating organization (South-Ukrainian Nuclear Power Plant) when extending its designated resource / service life, is shown. timely to clarify complex factors, technical aspects and parameters, as well as – their possible negative effects on the safe operation of systems and elements of nuclear power plants.
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Marcus, Gail H. "Nuclear Power after Fukushima." Mechanical Engineering 133, no. 12 (December 1, 2011): 27–29. http://dx.doi.org/10.1115/1.2011-dec-2.

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This article discusses advanced reactor technologies that are now getting renewed attention after the Fukushima nuclear plant accident. Interest in smaller reactors has been growing in recent years. Some of these designs have advantages over the traditional large light water reactors (LWRs) for certain applications. The smaller designs carry less of an inventory of nuclear material, so there is less material at risk in an accident involving a release. Proponents of small modular reactors (SMRs) point to cost savings due to the factory fabrication and shorter construction times. They have significant advantages for countries with small grids, where a current 1500 MWe reactor would exceed demand and threaten grid stability. Other designs that are getting the most attention at present are small or medium LWR concepts. In addition to their smaller size, these designs differ from current large, light-water designs in that most of them use an “integral” design. Most major reactor components are inside the reactor pressure vessel, thus significantly reducing the threat of a major loss-of-coolant accident.
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Dissertations / Theses on the topic "Nuclear reactor vessel"

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Everson, Matthew S. "The design of a reduced diameter Pebble Bed Modular Reactor for reactor pressure vessel transport by railcar." Thesis, Massachusetts Institute of Technology, 2009. http://hdl.handle.net/1721.1/53295.

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Thesis (S.M. and S.B.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2009.
Cataloged from PDF version of thesis.
Includes bibliographical references (p. 92).
Many desirable locations for Pebble Bed Modular Reactors are currently out of consideration as construction sites for current designs due to limitations on the mode of transportation for large RPVs. In particular, the PBMR-400 design developed by PBMR Pty of South Africa uses an RPV with an outer diameter of 6.4 meters. Since current SCHNABEL railcars can only haul components up to 4.3 meters wide, the only other possibility for transport is by barge, which limits construction to sites accessible by river, lake or coast. Designing a PBMR with a core able to fit within an RPV able to be transported by railcar would be extremely valuable, especially for potential inland sites only accessible by railway, such as those in the Canadian Oil Sands at which the PBMR would be utilized for oil extraction processes. Therefore, a study was conducted to determine the feasibility of a Pebble Bed Modular Reactor design operating at 250 MWth with a core restricted to fitting inside an RPV with an outer diameter of 4.3 meters. After reviewing the performance of various core configurations satisfying this constraint, an optimized PBMR design operating at this power was found. This new design uses the same fuel management scheme as the PBMR 400, as well as similar inlet and outlet coolant temperatures. This MPBR-250 design includes a pebble bed with an outer diameter of 2.7 meters, an outer reflector 50 cm thick and 12.5% enriched fuel. A mixture of graphite pebbles of 11.7% is also included in the pebble bed to produce an equilibrium core with minimal excess reactivity.
(cont.) This thesis shows that the MPBR-250 can perform up to the standards of the PBMR-400 design with respect to power peaking factors, peak temperatures and RPV fast fluences and can also increase fuel burnup to nearly 110 GWd/T. In addition, the MPBR-250 is a much more agile design, able to be deployed at a wider variety of locations because its RPV can be transported by railcar.
by Matthew S. Everson.
S.M.and S.B.
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Tanco, André. "Implementation of 3D-Imaging technique for visual testing in a nuclear reactor pressure vessel." Thesis, KTH, Maskinkonstruktion (Inst.), 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-157475.

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This master thesis has been performed by request of Dekra Industrial AB. Dekra Industrial AB is a Swedish subsidiary company of the German company Dekra and works for example with safety inspections within the nuclear power industry. The inspections performed by the company are often non-destructive testing (NDT) such as visual inspections of nuclear reactor pressure vessels. The inspection methods used today are considered to be further developed and there is a strong demand of improving the visual inspection. 3D-Imaging techniques are starting to be used as a measuring tool within the industry and could be a potential aid tool for the visual inspection. The purpose with the master thesis is to gain an understanding of 3D-Imaging technique to propose a suitable implementation so that it may be used as an aid tool for visual inspection. The main goal with the master thesis work is to gain knowledge about 3D-Imaging techniques and propose an implementation which may be used in the nuclear power industry. However there are different types of techniques and all of them have advantages and disadvantages. The method began with a comprehensive study about 3D-Imaging techniques, optics of 3D-Imaging and behaviour of electronics in radioactive environment. Information that could not be acquired by literature alone is acquired by interviews and meetings. The chosen 3D-Imaging technique that was considered to be the most suitable was structural light. Structural light is built on a triangulation principle that uses a projector and a camera for acquiring 3D coordinates. By using patterns displayed by the projector onto the object the camera may detect the reflected patterns and thus creating 3D coordinates. A structural light system was built and tested. The main test consisted about a two-level factorial design. The tested factors were triangulation angle, brightness and measurement distance. The test run that had the largest triangulation angle, highest brightness and shortest measurement distance gave the best accuracy. The accuracy was determined by measuring the flatness of the object. The best accuracy was measured to 91.5 μm. Besides the accuracy the technique has proven its potential by being able to scan weld tests and reconstruct well defined point clouds of the weld profiles. In conclusion the goal of the master thesis was reached and the demanded accuracy was reached. The accuracy is comparable with some industrial systems available today. This was possible due to use of a high resolution still camera. Since the camera and projector are commercially available products the tests proves that there is room for further improvements in order to reach better and a more robust accuracy. Keywords: Dekra Industrial AB, Visual testing, Imaging technique, Structural light
Detta examensarbete har utförts på uppdrag av Dekra Industrial AB. Dekra Industrial AB är ett dotterbolag till Dekra. Dekra Industrial AB arbetar främst med kontroller och provningar inom industrin. Kärnkraftindustrin är en industrigren där DEKRA arbetar med sådan kontroll Inspektionerna som utförs består huvudsakligen av oförstörande provning såsom visuell provning. Metoderna som används idag behöver vidareutvecklas och det finns en stark efterfrågan att förbättra den visuella inspektionen. 3D-avbildningsteknik är allt vanligare inom industrin idag och skulle kunna användas som ett mäthjälpmedel för att komplettera den visuella inspektionen. Syftet med examensarbetet är att få en förståelse för hur väl tekniken fungerar samt att föreslå en tillämpning där den kan komma att användas som ett komplement till den visuella inspektionen. Målet med arbetet är att ta fram underlag och föreslå en tillämpning för provning i högstrålande miljö. 3D-avbildningsteknik är ett generellt namn för många olika typer av tekniker som har sina fördelar respektive nackdelar. Arbetet inleds med en litteraturstudie kring 3D-avbildningstekniker, fysik med avseende på avbildningsteknik, den visuella proceduren idag samt hur elektronik påverkas av högstrålande miljö. Information som inte kan fås via studier inhämtas via intervjuer och möten. Tekniken som valdes att analyseras var strukturerat ljus. Tekniken bygger på en trianguleringsprincip som använder en projektor och kamera för att tillförskaffa 3D-koordinater. Genom att projicera mönster på ett objekt kan kameran detektera det reflekterade mönstret och på så vis skapa 3D koordinater. Ett strukturerat ljus system ställdes upp och testades. Testet bestod huvudsakligen av en försöksplanering där de testade faktorerna var trianguleringsvinkel, ljusstyrka och mätavstånd. Testuppställningen som gav bäst resultat var med störst trianguleringsvinkel, högsta ljusstyrka samt kortast mätavstånd. Noggrannheten bestämdes genom att mäta planheten på objektet. Den bästa noggrannheten som uppnåddes med testet var 91.5 μm. Förutom den goda noggrannheten har tekniken visat sin potential genom att avbilda ett svetsprov som genererade ett väldefinierat punktmoln av svetsprofilen. Sammanfattningsvis uppfylldes målen och det uppställda systemet gav en noggrannhet som är jämförbar med en del system ute på marknaden. Detta var möjligt på grund av att en högupplöst stillbildskamera användes. Det finns potential för förbättringar då komponenterna som används i systemet är kommersiella produkter. Nyckelord: Dekra Industrial AB, Visuell inspektion, Avbildningsteknik, Strukturerat ljus
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Buongiorno, Jacopo 1971. "Conceptual design of a lead-bismuth cooled fast reactor with in-vessel direct-contact steam generation." Thesis, Massachusetts Institute of Technology, 2001. http://hdl.handle.net/1721.1/32205.

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Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 2001.
Includes bibliographical references (p. 357-366).
The feasibility of a lead-bismuth (Pb-Bi) cooled fast reactor that eliminates the need for steam generators and coolant pumps was explored. The working steam is generated by direct contact vaporization of water and liquid metal in the chimney above the core and then is sent to the turbine. The presence of a lighter fluid in the chimney drives the natural circulation of the Pb- Bi within the reactor pool. Three key technical issues were addressed: 1) the maximum thermal power removable by direct contact heat transfer without violating the fuel, clad and vessel temperature limits, 2) the consequences of Pb-Bi aerosol transport on the design and operation of the turbine and 3) the release of radioactive polonium (a product of coolant activation) to the steam. Modeling of the multi-phase phenomena occurring in the chimney confirmed the effectiveness of the direct contact heat transfer mode within a well-defined design envelope for the reactor power, chimney height and steam superheat. A 1260MWth power is found possible for 10m chimney height and 25°C superheat. The temperature of the low-nickel steel clad is maintained below 600°C, which results in limited corrosion if tight control of the coolant oxygen concentration is adopted.
Generation, transport and deposition of Pb-Bi aerosols were also modeled. It was found that the design of a chevron steam separator reduces the heavy liquid metal in the steam lines by about three orders of magnitude. Nevertheless, the residual Pb-Bi is predicted to cause embrittlement of the turbine blades. Four solutions to this problem were assessed: blade coating, employment of alternative materials, electrostatic precipitation and oxidation of the Pb-Bi droplets. An experimental campaign was conducted to investigate the polonium release from a hot Pb- Bi bath to a gas-streamn. Th thermodynamics of the polonium hydride formation reaction (free- energy vs. temperature). as welQ as the vapor pressure of the lead-polonide were measured and then utilized to model the polonium transport in the reactor. It was found that the polonium concentration in the steam and on the surface of the power cycle components is significantly above the acceptable limits, which makes the very concept of a direct contact reactor open to question.
by Jacopo Buongiorno.
Ph.D.
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Viehrig, Hans-Werner, Eberhard Altstadt, Mario Houska, Gudrun Mueller, Andreas Ulbricht, Joerg Konheiser, and Matti Valo. "Investigation of decommissioned reactor pressure vessels of the nuclear power plant Greifswald." Helmholtz-Zentrum Dresden - Rossendorf, 2018. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-235681.

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The investigation of reactor pressure vessel (RPV) material from the decommissioned Greifswald nuclear power plant representing the first generation of Russian-type WWER-440/V-230 reactors offers the opportunity to evaluate the real toughness response. The Greifswald RPVs of 4 units represent different material conditions as follows: • Irradiated (Unit 4), • irradiated and recovery annealed (Units 2 and 3), and • irradiated, recovery annealed and re-irradiated (Unit1). The recovery annealing of the RPV was performed at a temperature of 475° for about 152 hours and included a region covering ±0.70 m above and below the core beltline welding seam. Material samples of a diameter of 119 mm called trepans were extracted from the RPV walls. The research program is focused on the characterisation of the RPV steels (base and weld metal) across the thickness of the RPV wall. This report presents test results measured on the trepans from the beltline welding seam No. SN0.1.4. and forged base metal ring No. 0.3.1. of the Units 1 2 and 4 RPVs. The key part of the testing is focussed on the determination of the reference temperature T0 of the Master Curve (MC) approach following the ASTM standard E1921 to determine the facture toughness, and how it degrades under neutron irradiation and is recovered by thermal annealing. Other than that the mentioned test results include Charpy-V and tensile test results. Following results have been determined: • The mitigation of the neutron embrittlement of the weld and base metal by recovery annealing could be confirmed. • KJc values of the weld metals generally followed the course of the MC though with a large scatter. • There was a large variation in the T0 values evaluated across the thickness of the multilayered welding seams. • The T0 measured on T-S oriented SE(B) specimens from different thickness locations of the welding seams strongly depended on the intrinsic structure along the crack front. • The reference temperature RT0 determined according to the “Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs - VERLIFE” and the fracture toughness lower bound curve based thereon are applicable on the investigated weld metals. • A strong scatter of the fracture toughness KJc values of the recovery annealed and re-irradiated and the irradiated base metal of Unit 1 and 4, respectively is observed with clearly more than 2% of the values below the MC for 2% fracture probability. The application of the multimodal MC-based approach was more suitable and described the temperature dependence of the KJc values in a satisfactory manner. • It was demonstrated that T0 evaluated according to the SINTAP MC extension represented the brittle fraction of the data sets and is therefore suitable for the nonhomogeneous base metal. • The efficiency of the large-scale thermal annealing of the Greifswald WWER 440/V230 Unit 1 and 2 RPVs could be confirmed.
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Maples, Allen B. "Design of a robust acoustic positioning system for an underwater nuclear reactor vessel inspection robot." Thesis, This resource online, 1993. http://scholar.lib.vt.edu/theses/available/etd-06232009-063217/.

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Wells, Peter Benjamin. "The Character, Stability and Consequences of Mn-Ni-Si Precipitates in Irradiated Reactor Pressure Vessel Steels." Thesis, University of California, Santa Barbara, 2016. http://pqdtopen.proquest.com/#viewpdf?dispub=10103547.

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Formation of a high density of Mn-Ni-Si nanoscale precipitates in irradiated reactor pressure vessel steels could lead to severe, unexpected embrittlement, which may limit the lifetimes of our nation’s light water reactors. While the existence of these precipitates was hypothesized over 20 years ago, they are currently not included in embrittlement prediction models used by the Nuclear Regulatory Commission. This work aims to investigate the mechanisms and variables that control Mn-Ni-Si precipitate (MNSP) formation as well as correlate their formation with hardening and embrittlement.

A series of RPV model steels with systematic variations in Cu and Ni contents, two variables that have been shown to have a dominant effect on hardening, were irradiated in a series of test reactor and power reactor surveillance irradiations. Atom probe tomography (APT) measurements show that large volume fractions (fv) of MNSPs form in all the steels irradiated at high fluence, even those containing no added Cu, which were previously believed to have low sensitivity to embrittlement. It is demonstrated that while Cu enhances the rate of MNSP formation, it does not appear to significantly alter their saturation fv or composition. The high fluence MNSPs have compositions consistent with known intermetallic phases in the Mn-Ni-Si system and have fv very near those predicted by equilibrium thermodynamic models. In addition, X-ray diffraction experiments by collaborators shows that these precipitates also have the expected crystal structure of the predicted Mn-Ni-Si phases.

Post irradiation annealing experiments are used to measure the hardness recovery at various temperatures as well as to determine if the large f v of MNSPs that form under high fluence neutron irradiation are thermodynamically stable phases or non-equilibrium solute clusters, enhanced or induced by irradiation, respectively. Notably, while post irradiation annealing of a Cu-free, high Ni steel at 425°C results in dissolution of most precipitates, a few larger MNSPs appear to remain stable and may begin to coarsen after long times. A cluster dynamics model rationalizes the dissolution and reduction in precipitate number density, since most are less than the critical radius at the annealing temperature and decomposed matrix composition. The stability of larger precipitates suggests that they are an equilibrium phase, consistent with thermodynamic models.

Charged particle irradiations using Fe3+ ions are also used to investigate the precipitates which form under irradiation. Two steels irradiated to a dose of 0.2 dpa using both neutrons and ions show precipitates with very similar compositions. The ion irradiation shows a smaller f v, likely due to the much higher dose rate, which has been previously shown to delay precipitation to higher fluences. While the precipitates in the ion irradiated condition are slightly deficient in Mn and enriched in Ni and Si compared to neutron irradiated condition, the overall similarities between the two conditions suggest that ion irradiations can be a very useful tool to study the susceptibility of a given steel to irradiation embrittlement.

Finally, the large fv of MNSPs that are shown to form in all steels, including those low in Cu, at high fluence, even those without added Cu, result in large amounts of hardening and embrittlement. A preliminary embrittlement prediction model, which incorporates MNSPs at high fluence, is presented, along with results from a recent test reactor irradiation to fluences representative of extended lifetimes. This model shows very good agreement with the data.

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Petersson, Jens. "CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel." Thesis, Linköpings universitet, Mekanisk värmeteori och strömningslära, 2014. http://urn.kb.se/resolve?urn=urn:nbn:se:liu:diva-112796.

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In this work a cooling system connected to a reactor pressure vessel has been studied using the CFD method for the purpose of investigating the strengths and shortcomings of using CFD as a tool in similar fluid flow problems within nuclear power plants. The cooling system is used to transport water of 288K (15°C) into a nuclear reactor vessel filled with water of about 555K (282°C) during certain operating scenarios. After the system has been used, the warm water inside the vessel will be carried into the cooling system by buoyancy forces. It was of interest to investigate how quickly the warm water moves into the cooling system and how the temperature field of the water changes over time. Using the open source CFD code OpenFOAM 2.3.x and the LES turbulence modelling method, a certain operating scenario of the cooling system was simulated. A simplified computational domain was created to represent the geometries of the downcomer region within the reactor pressure vessel and the pipe structure of the cooling system. Boundary conditions and other domain properties were chosen and motivated to represent the real scenario as good as possible. For the geometry, four computational grids of different sizes and design were generated. Three of these were generated using the ANSA pre-processing tool, and they all have the same general structure only with different cell sizes. The fourth grid was made by the OpenFOAM application snappyHexMesh, which automatically creates the volume mesh with little user input. It was found that for the case at hand, the different computational grids produced roughly the same results despite the number of cells ranging from 0,14M to 3,2M. A major difference between the simulations was the maximum size of the time steps which ranged from 0,3ms for the finest ANSA mesh to 2ms for the snappy mesh, a difference which has a large impact on the total time consumption of the simulations. Furthermore, a comparison of the CFD results was made with those of a simpler 1D thermal hydraulic code, Relap5. The difference in time consumption between the two analyses were of course large and it was found that although the CFD analysis provided more detailed information about the flow field, the cheaper 1D analysis managed to capture the important phenomena for this particular case. However, it cannot be guaranteed that the 1D analysis is sufficient for all similar flow scenarios as it may not always be able to sufficiently capture phenomena such as thermal shocks and sharp temperature gradients in the fluid. Regardless of whether the CFD method or a simpler analysis is used, conservativeness in the flow simulation results needs to be ensured. If the simplifications introduced in the computational models cannot be proved to always give conservative results, the final simulation results need to be modified to ensure conservativeness although no such modifications were made in this work.
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Longmire, Pamela. "Nonparametric statistical methods applied to the final status decommissioning survey of Fort St. Vrains prestressed concrete reactor vessel." The Ohio State University, 1998. http://rave.ohiolink.edu/etdc/view?acc_num=osu1407398430.

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Lim, Joven Jun Hua. "Electron microscopy studies of precipitation in nuclear reactor pressure vessel steels under neutron irradiation and thermally ageing." Thesis, University of Oxford, 2014. http://ora.ox.ac.uk/objects/uuid:33daab2c-5c3f-466b-bdd6-0cc022169a6b.

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Maintaining the safe operation of nuclear power plants (NPPs) is crucial. This requires fully understanding the mechanism of long term irradiation and thermal ageing, as well as their effects, on components including the reactor pressure vessel (RPV). The research community is collecting data that will be required to support the case for extending the operation of western-type NPPs beyond that of 60 years. One of the current dilemmas faced by the long-term operation of RPVs is the formation of nanometre scale precipitates. These precipitates are known to cause embrittlement where it increases the ductile-to-brittle transition temperature of the RPV steels. The chemistry of these precipitates is strongly dependent on the chemistry of the RPV steels. In general, these precipitates can be categorised into two types, copper-rich precipitates (CRPs) and manganese-nickel (-enriched) precipitates (MNPs) [1, 2]. The concentration of copper in the precipitates depends on the bulk content of the steel [3]. The formation mechanism of the precipitates under neutron irradiation and thermal ageing, and their influence on material degradation at high neutron fluence (Φt), is still unclear. To understand the long term precipitation under irradiation and thermal ageing, high nickel and copper containing RPV steels with a similar microstructure an chemical composition as those currently in service were subjected to either neutron irradiation (to high neutron fluences, Φt ≥ 5 x 1023 neutrons.m-2) or thermal ageing (for as long as ≈ 50,000 hours). CRPs and MNPs were both detected. The co-precipitation of the CRPs and MNPs were observed in thermally aged steels. The development of crystal structures in the CRPs is believed to be dependent on the size of the precipitates and the ambient temperature. When the CRPs reached a critical size, they underwent the martensitic transformation from BCC→9R→3R→FCC or FCT. The CRPs preferentially nucleate heterogeneously at the dislocation lines. Chemical analysis suggests that most of the CRPs are iron free. Under thermal ageing, the MNPs were found to precipitate at the interface of the CRPs and the matrix. These MNPs are found to be iron free too. Larger MNPs were often found to be at CPRs that were associated with dislocation lines. Also, based on the volume fraction observed, it is possible to suggest that the kinetics of nucleation and growth of the MNPs are relatively slow compared to the CRPs. This is in good agreement with the simulations reported in Refs. [4, 5]. It is the first time the MNPs are directly imaged from neutron irradiation low copper steels using electron microscopy. These irradiation-induced MNPs are densely populated in the neutron irradiated samples. It was found that the irradiation-induced MNPs are more sensitive to electron beams. It was thought that this was due to a relatively large amount of point defects present in the irradiation-induced MNPs.
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Ruan, Xiaoyong. "Structural Integrity Assessment of Nuclear Energy Systems." Kyoto University, 2020. http://hdl.handle.net/2433/253517.

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Books on the topic "Nuclear reactor vessel"

1

Server, William L., and Milan Brumovský, eds. International Review of Nuclear Reactor Pressure Vessel Surveillance Programs. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2018. http://dx.doi.org/10.1520/stp1603-eb.

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Johnson, R. E. Radiation effects on reactor pressure vessel supports. Washington, DC: U.S. Nuclear Regulatory Commission, 1996.

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Johnson, R. Radiation effects on reactor pressure vessel supports. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1996.

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Hawthorne, J. R. Accelerated irradiation test of Gundremmingen reactor vessel trepan material. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

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Croneberg, August W. In-vessel zircaloy oxidation/hydrogen generation behavior during severe accidents. Washington, D.C: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

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Croneberg, August W. In-vessel zircaloy oxidation/hydrogen generation behavior during severe accidents. Washington, D.C: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1990.

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McCabe, Donald E. Fracture evaluation of surface cracks embedded in reactor vessel cladding. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

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U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering., University of Tennessee Knoxville, and Oak Ridge National Laboratory, eds. Extrapolation of the J-R curve for predicting reactor vessel integrity. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

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Remec, I. Neutron spectra at different high flux isotope reactor (HFIR) pressure vessel surveillance locations. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

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Remec, I. Neutron spectra at different high flux isotope reactor (HFIR) pressure vessel surveillance locations. Washington, DC: Division of Safety Issue Resolution, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

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Book chapters on the topic "Nuclear reactor vessel"

1

Moore, Kenneth E., A. S. Heller, and Arthur L. Lowe. "Chemical Composition of Nuclear Reactor Vessel Welds." In Effects of Radiation on Materials: 12th International Symposium Volume II, 1046–58. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 1985. http://dx.doi.org/10.1520/stp87019850030.

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Gérard, Robert, and Rachid Chaouadi. "Reactor Pressure Vessel Surveillance Programs in Belgium." In International Review of Nuclear Reactor Pressure Vessel Surveillance Programs, 250–75. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2018. http://dx.doi.org/10.1520/stp160320170001.

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Slugeň, V. "Microstructural Analysis of Nuclear Reactor Pressure Vessel Steels." In Mössbauer Spectroscopy in Materials Science, 119–30. Dordrecht: Springer Netherlands, 1999. http://dx.doi.org/10.1007/978-94-011-4548-0_12.

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Ošmera, B., and M. Holman. "Integral Experiments for Reactor Pressure Vessel Neutron Exposure Evaluation." In Nuclear Data for Science and Technology, 650–52. Berlin, Heidelberg: Springer Berlin Heidelberg, 1992. http://dx.doi.org/10.1007/978-3-642-58113-7_185.

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Duo, J. I., J. Chen, J. A. Kulesza, A. H. Fero, C. S. Yoo, and B. C. Kim. "Korean Standard Nuclear Plant Ex-Vessel Neutron Dosimetry Program Ulchin 4." In Reactor Dosimetry: 14th International Symposium, 13–21. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2012. http://dx.doi.org/10.1520/stp49599t.

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Duo, J. I., J. Chen, J. A. Kulesza, A. H. Fero, C. S. Yoo, and B. C. Kim. "Korean Standard Nuclear Plant Ex-Vessel Neutron Dosimetry Program Ulchin 4." In Reactor Dosimetry: 14th International Symposium, 13–21. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2012. http://dx.doi.org/10.1520/stp155020120002.

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Xu, Junying, Lei Zhang, Dekui Zhan, Huiyong Zhang, Yahelle Laroche, Hui Guo, and Guillaume Niessen. "Study of Potential for In-Vessel Retention Through External Reactor Vessel Flooding: Code Comparison." In Proceedings of The 20th Pacific Basin Nuclear Conference, 601–15. Singapore: Springer Singapore, 2017. http://dx.doi.org/10.1007/978-981-10-2311-8_56.

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Singh, Upendra, Vivek Shrivastav, and Rabindranath Sen. "Life Estimation Strategy for a Nuclear Reactor Pressure Vessel." In Lecture Notes in Mechanical Engineering, 31–51. Singapore: Springer Singapore, 2020. http://dx.doi.org/10.1007/978-981-15-4779-9_4.

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Brumovský, Milan. "The Bases for WWER Vessel Surveillance Programs: General Requirements." In International Review of Nuclear Reactor Pressure Vessel Surveillance Programs, 54–67. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2018. http://dx.doi.org/10.1520/stp160320160161.

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Yoo, Choon Sung, and Byoung Chul Kim. "Neutron Flux Reduction Programs for Reactor Pressure Vessel of Korea Nuclear Unit 1." In Reactor Dosimetry: 14th International Symposium, 249–63. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959: ASTM International, 2012. http://dx.doi.org/10.1520/stp49617t.

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Conference papers on the topic "Nuclear reactor vessel"

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Aquaro, D., M. D. Carelli, G. Forasassi, R. Lo Frano, and N. Zaccari. "Seismic Response of Reactor Vessel Internals in the IRIS Reactor." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89579.

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The intent of this paper is the presentation and discussion of a methodology for the evaluation and analysis of seismic loads effects on a nuclear power plant. To help in focussing the presented methodology, a preliminary simplified analysis of an integral, medium size next generation PWR reactor structure (IRIS project, an integral configuration PWR under study by an international group) was considered as an application example also for models/codes evaluation. The performed preliminary seismic analysis, even though by no means complete, is intended to evaluate the method of calculating the effects of dynamic loads propagation to the reactor internals for structural design as well as geometrical and functional optimisation purposes. To this goal, finite element method and separated (sub) structures approaches were employed for studying the overall dynamic behaviour of the nuclear reactor vessel. The analysis was set up by means of numerical models, implemented on the MARC FEM code, on the basis of Design Response Spectra as indicated on the relevant rules for Nuclear Power Plants (NRC 1.60) design. The seismic analysis is indented to evaluate the dynamic loads propagated from the ground through the Containment System and Vessel to the Steam Generator’s tubes.
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Li, Fei, and Mohammad Modarres. "Uncertainty Characterization of Reactor Vessel Fracture Toughness." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22647.

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To perform fracture mechanics analysis of reactor vessel, fracture toughness (KIc) at various temperatures would be necessary. In a best estimate approach, KIc uncertainties resulting from both lack of sufficient knowledge and randomness in some of the variables of KIc must be characterized. Although it may be argued that there is only one type of uncertainty, which is lack of perfect knowledge about the subject under study, as a matter of practice KIc uncertainties can be divided into two types: aleatory and epistemic. Aleatory uncertainty is related to uncertainty that is very difficult to reduce, if not impossible; epistemic uncertainty, on the other hand, can be practically reduced. Distinction between aleatory and epistemic uncertainties facilitates decision-making under uncertainty and allows for proper propagation of uncertainties in the computation process. Typically, epistemic uncertainties representing, for example, parameters of a model are sampled (to generate a “snapshot,” single-value of the parameters), but the totality of aleatory uncertainties is carried through the calculation as available. In this paper a description of an approach to account for these two types of uncertainties associated with KIc has been provided.
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Li, Guoyun, Yukun Wu, Guofu Jiang, Juan Huang, and Haisheng Zhang. "Irradiation-Embrittlement of Reactor Pressure Vessel Steel." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29240.

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Irradiated for 4 cycles inside the reactor, the irradiation surveillance capsule (ISC) was withdrawn and transported to Nuclear Power Institute of China (NPIC) for dismantling in the hot cell. This paper presented tensile tests and Charpy-V impact tests on non-irradiated and irradiated specimens. Tensile tests were performed at room temperature and 300 °C respectively, and Charpy-V impact tests were performed at series temperatures. Instrumented impact curves and tensile curves were obtained and analyzed. Impact absorbed energy, lateral expansion and crystallinity transition curves were plotted by a hyperbolic tangent function. The irradiation-induced shift of transition temperature, upper shelf energy, tensile strength and yield strength were determined. And irradiation-embrittlement effect on RPV was evaluated by prediction and measurement of irradiation-induced transition temperature shift. Results showed that the parameters including neutron fluence and transition temperature were in normal range, and the next capsule could be withdrawn according to original irradiation monitoring program.
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Tagawa, Akihiro, Masashi Ueda, and Takuya Yamashita. "Development of the ISI Device for Fast Breeder Reactor MONJU Reactor Vessel." In 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89230.

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In-service inspection (ISI) is carried out to confirm the integrity of the main components of the Fast Breeder Reactor (FBR) “MONJU”. The weld-joints are examined by using an inspection device which has a glass fiber scope for visual examination and a horizontally polarized shear (SH) wave electromagnetic acoustic transducer (EMAT) for volumetric testing. The ambient temperature during the inspection is 200°C and the irradiation field is 10 Sv/hr. A new inspection device has been developed in order to improve the visual test performance, volumetric test performance and controllability of the inspection device reflecting the experience of the original test. In this paper, detail of the new inspection device and the test results of sensors such as the CCD camera, EMAT and bead sensor are reported. The paper also reports on the CCD camera cooling system and other components.
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Takamatsu, Misao, Kazuyuki Imaizumi, Akinori Nagai, Takashi Sekine, and Yukimoto Maeda. "Development of Observation Techniques in Reactor Vessel of Experimental Fast Reactor Joyo." In 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75088.

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During the investigation of an incident that occurred with the experimental fast reactor Joyo, In-Vessel Observations (IVO) using a standard Video Camera (VC) and a Radiation-Resistant Fiberscope (RRF) took place at (1) the top of the Sub-Assemblies (S/As) and the In-Vessel Storage rack (IVS), (2) the bottom face of the Upper Core Structure (UCS) under the condition with the level of sodium at −50 mm below the top of the S/As. A simple 6 m overhead view of each S/A, through the fuel handling or inspection holes etc, was photographed using a VC fixed to the Rotating-Plug (R/P) with the acrylic panel for making observations of the top of S/As and IVS. About 650 photographs were required to create a composite photograph of the top of the entire S/As and IVS, and a resolution was estimated to be approximately 1mm. In order to observe the bottom face of the UCS, a Remote Handling Device (RHD) equipped with RRFs (approximately 13 m long) was specifically developed for Joyo with a tip that could be bent into an L-shape and inserted into the 70 mm gap between the top of the S/As and the bottom of the UCS. A total of about 35,000 photographs were needed for the full investigation. Regarding the resolution, the sodium flow regulating grid of 0.8mm in thickness could be discriminated, and the base of thermocouple sleeves 6 mm in diameter located 450 mm above the top of S/As were also clearly observed. In both types of observations, it was confirmed that lighting adjustments play a critical role. Particularly in narrow space observations, scattered lighting with automatic dimming controlled light source was available for achieving close observations of the in-vessel structures. In addition to the successful result of the incident investigation, these experiments provided valuable insights for use in further improving and verifying in-vessel observation techniques in Sodium cooled Fast Reactors (SFRs).
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Ide, Hiroshi, Akihiro Kimura, Hiroshi Miura, Yoshiharu Nagao, Naohiko Hori, and Masanori Kaminaga. "Investigation on Integrity of JMTR Reactor Pressure Vessel." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30238.

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Visual observation of inner side of a reactor pressure vessel of Japan Materials Testing Reactor (JMTR) was carried out using an underwater camera before the JMTR refurbishment work from the view point of its long term utilization, because the reactor pressure vessel of the JMTR will be used continuously after restart of the JMTR. As a result of the visual observation, the harmful wound was not confirmed. Moreover, there was no loosening of the bolts and the screws. On the other hand, adhesion materials which can be easily removed using the gauze were observed around nozzles in a top closure of the reactor pressure vessel. A major component of the adhesion materials is an iron as a result of the componential analysis. However, no significant problem affecting the integrity of the reactor pressure vessel was observed, and then the integrity of the reactor pressure vessel was confirmed. From view points of the stress corrosion cracking, fast neutron fluence and fatigue, it became clear that the reactor pressure vessel of the JMTR can be used for more than 20 years. The visual observation by the underwater camera is to be carried out periodically to confirm the integrity of the reactor pressure vessel in future.
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Zhang, Xiaojun, Jianhua Zhang, Jie Yuan, and Manhong Li. "Development of an underwater robot for nuclear reactor vessel." In 2013 IEEE International Conference on Robotics and Biomimetics (ROBIO). IEEE, 2013. http://dx.doi.org/10.1109/robio.2013.6739712.

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Kujawski, J. M., D. M. Kitch, and L. E. Conway. "The IRIS Spool-Type Reactor Coolant Pump." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22572.

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IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. One of the innovative features of the IRIS design is the adoption of a reactor coolant pump (called “spool” pump) which is completely contained inside the reactor vessel. Background, status and future developments of the IRIS spool pump are presented in this paper.
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Huang, Kuan-Rong, Chin-Cheng Huang, and Hsoung-Wei Chou. "Probabilistic Fracture Mechanics Analysis for Degraded Reactor Pressure Vessel in Pressurized Water Reactor Nuclear Power Plant." In ASME 2014 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/pvp2014-28595.

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Cumulative radiation embrittlement is one of the main causes for the degradation of PWR reactor pressure vessels over their long term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the United States is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Further, two overcooling transients of steam generator tube rupture and pressurized thermal shock addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR reactor pressure vessels and can be also referred as its regulatory basis in Taiwan.
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Boggess, Cheryl L., Bruce A. Bishop, Nathan A. Palm, and Owen F. Hedden. "Risk-Informed Pressurized Water Reactor Vessel Inspection Interval Extension." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49429.

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The methodology discussed in this paper provides a risk informed basis for decreasing the frequency of inspection for the Pressurized Water Reactor (PWR) reactor pressure vessel (RPV). The decrease in frequency is based on extending the interval between inspections from the current interval of 10 years to 20 years. Results of pilot studies on typical designs of PWR vessels show that the change in risk associated with extending the inspection interval by more than 10 years is within the guidelines specified in U.S. Regulatory Guide 1.174 for insignificant change in risk. The current requirements for inspection of reactor vessel pressure-containing welds have been in effect since the 1989 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, supplemented by U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.150, June 1981. The manner in which these examinations are conducted has recently been augmented by Appendix VIII of Section XI, 1996 Addenda, as implemented by the NRC in amendment to 10CFR50.55a effective November 22, 1999. This paper summarizes the insignificant change in risk results for the PWR pilot-plant studies, including the effects of fatigue crack growth and in-service inspection of postulated surface-breaking flaws. These results demonstrate that the proposed RPV inspection interval extension is a viable option for the industry.
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Reports on the topic "Nuclear reactor vessel"

1

Love, E. F., K. A. Pauley, and B. D. Reid. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors. Office of Scientific and Technical Information (OSTI), September 1995. http://dx.doi.org/10.2172/130639.

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Natesan, K., S. Majumdar, P. S. Shankar, and V. N. Shah. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel. Office of Scientific and Technical Information (OSTI), March 2007. http://dx.doi.org/10.2172/925328.

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J. K. Wright and R. N. Wright. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803). Office of Scientific and Technical Information (OSTI), July 2010. http://dx.doi.org/10.2172/989891.

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J. K. Wright and R. N. Wright. Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803). Office of Scientific and Technical Information (OSTI), April 2008. http://dx.doi.org/10.2172/952022.

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Acton, R. U., W. Gill, D. J. Sais, D. H. Schulze, and J. T. Nakos. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing. Office of Scientific and Technical Information (OSTI), May 1996. http://dx.doi.org/10.2172/274130.

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Nanstad, Randy, Mikhail Sokolov, and William Server. Preliminary Plan for Evaluation of Reactor Pressure Vessel Surveillance Materials from Palisades Nuclear Generating Station. Office of Scientific and Technical Information (OSTI), February 2020. http://dx.doi.org/10.2172/1770673.

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Nakos, J. T., S. T. Rosinski, and R. U. Acton. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section. Office of Scientific and Technical Information (OSTI), November 1994. http://dx.doi.org/10.2172/10106584.

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Searfass, Clifford T., Owen M. Malinowski, and Jason K. Van Velsor. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components. Office of Scientific and Technical Information (OSTI), March 2015. http://dx.doi.org/10.2172/1173231.

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Ren, Weiju, and Totemeier Terry. Assessment of Negligible Creep, Off-Normal Welding and Heat Treatment of Gr91 Steel for Nuclear Reactor Pressure Vessel Application. Office of Scientific and Technical Information (OSTI), October 2006. http://dx.doi.org/10.2172/1093013.

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Pareige, P., K. F. Russell, R. E. Stoller, and M. K. Miller. Influence of long-term thermal aging on the microstructural evolution of nuclear reactor pressure vessel materials: An atom probe study. Office of Scientific and Technical Information (OSTI), March 1998. http://dx.doi.org/10.2172/573374.

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