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Статті в журналах з теми "Nuclear reactor vessel"
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.
Повний текст джерелаKondylakis, 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.
Повний текст джерелаKantsedalov, 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.
Повний текст джерелаKramskoi, 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.
Повний текст джерелаZabusov, 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.
Повний текст джерелаDombrovskii, 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.
Повний текст джерелаRosinski, 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.
Повний текст джерелаAzhagarason, 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.
Повний текст джерелаPopov, 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.
Повний текст джерелаMarcus, 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.
Повний текст джерелаДисертації з теми "Nuclear reactor vessel"
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.
Повний текст джерелаCataloged 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.
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.
Повний текст джерелаDetta 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
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.
Повний текст джерелаIncludes 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.
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.
Повний текст джерелаMaples, 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/.
Повний текст джерелаWells, 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.
Повний текст джерелаFormation 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.
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.
Повний текст джерелаLongmire, 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.
Повний текст джерелаLim, 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.
Повний текст джерелаRuan, Xiaoyong. "Structural Integrity Assessment of Nuclear Energy Systems." Kyoto University, 2020. http://hdl.handle.net/2433/253517.
Повний текст джерелаКниги з теми "Nuclear reactor vessel"
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.
Повний текст джерелаJohnson, R. E. Radiation effects on reactor pressure vessel supports. Washington, DC: U.S. Nuclear Regulatory Commission, 1996.
Знайти повний текст джерелаJohnson, 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.
Знайти повний текст джерелаHawthorne, 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.
Знайти повний текст джерелаCroneberg, 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.
Знайти повний текст джерелаCroneberg, 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.
Знайти повний текст джерелаMcCabe, 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.
Знайти повний текст джерелаU.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.
Знайти повний текст джерелаRemec, 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.
Знайти повний текст джерелаRemec, 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.
Знайти повний текст джерелаЧастини книг з теми "Nuclear reactor vessel"
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.
Повний текст джерелаGérard, 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.
Повний текст джерелаSlugeň, 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.
Повний текст джерелаOšmera, 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.
Повний текст джерелаDuo, 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.
Повний текст джерелаDuo, 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.
Повний текст джерелаXu, 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.
Повний текст джерелаSingh, 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.
Повний текст джерелаBrumovský, 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.
Повний текст джерелаYoo, 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.
Повний текст джерелаТези доповідей конференцій з теми "Nuclear reactor vessel"
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.
Повний текст джерелаLi, 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.
Повний текст джерелаLi, 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.
Повний текст джерелаTagawa, 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.
Повний текст джерелаTakamatsu, 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.
Повний текст джерелаIde, 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.
Повний текст джерелаZhang, 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.
Повний текст джерелаKujawski, 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.
Повний текст джерелаHuang, 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.
Повний текст джерелаBoggess, 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.
Повний текст джерелаЗвіти організацій з теми "Nuclear reactor vessel"
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.
Повний текст джерелаNatesan, 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.
Повний текст джерелаJ. 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.
Повний текст джерелаJ. 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.
Повний текст джерелаActon, 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.
Повний текст джерелаNanstad, 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.
Повний текст джерелаNakos, 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.
Повний текст джерелаSearfass, 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.
Повний текст джерелаRen, 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.
Повний текст джерелаPareige, 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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