Journal articles: 'Nuclear reactor vessel'

1

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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2

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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3

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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4

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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7

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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8

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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9

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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10

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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11

Langston, Lee S. "PBMR-A Future Failsafe Gas Turbine Nuclear Power Plant?" Mechanical Engineering 133, no. 08 (August 1, 2011): 54–59. http://dx.doi.org/10.1115/1.2011-aug-5.

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This article presents an overview of a pebble bed modular reactor (PBMR) power plant. A PBMR power plant is a gas turbine nuclear power plant that completely eliminates the possibility of a devastating loss-of-coolant accident. In a PBMR power plant, uranium dioxide nuclear fuel, coated with mass diffusion and radioactive fission product containment layers of pyrolytic carbon and silicon carbide, is formed into nuclear poppy seed-sized fuel particles. Some 15,000 of these are embedded in a tennis ball-sized graphite sphere, which is encased in a thin carbon shell, sintered, annealed and machined to a uniformed diameter of 6 cm. The PBMR reactor vessel, 90 ft high and 20 ft wide, is packed with about 450,000 heat-producing nuclear pebbles. Helium gas coolant then flows around and between the pebbles stacked in the reactor vessel, emerging at about 900°F. The Chinese are currently building two pebble reactors that will be used to generate steam for a conventional Rankine cycle.

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12

Wen, Lijing, Chao Guo, Tieping Li, and Chunming Zhang. "Stress Analysis for Reactor Coolant Pump Nozzle of Nuclear Reactor Pressure Vessel." Journal of Applied Mathematics and Physics 01, no. 06 (2013): 62–64. http://dx.doi.org/10.4236/jamp.2013.16013.

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13

Weng, Yu, Haitao Wang, Benan Cai, Hongfang Gu, and Haijun Wang. "Flow mixing and heat transfer in nuclear reactor vessel with direct vessel injection." Applied Thermal Engineering 125 (October 2017): 617–32. http://dx.doi.org/10.1016/j.applthermaleng.2017.07.040.

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14

Kuroda, Toshio, and Fukuhisa Matsuda. "Irradiation Embrittlement for Weldment of Nuclear Reactor Vessel Steel." Journal of the Japan Welding Society 65, no. 7 (1996): 568–75. http://dx.doi.org/10.2207/qjjws1943.65.7_568.

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15

Bibel, George. "Significant wastage of a nuclear reactor pressure vessel head." International Journal of Forensic Engineering 2, no. 1 (2014): 1. http://dx.doi.org/10.1504/ijfe.2014.059243.

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16

Horiya, T., T. Takeda, and K. Yamato. "Study on Underclad Cracking in Nuclear Reactor Vessel Steels." Journal of Pressure Vessel Technology 107, no. 1 (February 1, 1985): 30–35. http://dx.doi.org/10.1115/1.3264400.

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Susceptibility to underclad cracking in nuclear reactor vessel steels, such as SA533 Grade B Class 1 and SA508 Class 2, was studied in detail. A convenient simulation test method using simulated HAZ specimens of small size has been developed for quantitative evaluation of susceptibility to underclad cracks. The method can predict precisely the cracking behavior in weldments of steels with relative low crack susceptibility. The effect of chemical compositions on susceptibility to the cracking was examined systematically using the developed simulation test method and the following index was obtained from the test results: U=20[V]+7[C]+4[Mo]+[Cr]+[Cu]−0.5[Mn]+1.5log[X]X=Al…..Al/2N≦1X=2N…..Al/2N>1 It was confirmed that the new index (U) is useful for the prediction of crack susceptibility of the nuclear vessel steels; i.e., no crack initiation is detected in weldments in the roller bend test for steels having U value below 0.90.

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17

van Hoorebeke, L., J. van de Velde, D. Segers, L. Dorikens-Vanpraet, and S. Simula. "Positron Annihilation Measurements on Nuclear Reactor Pressure Vessel Steels." Le Journal de Physique IV 05, no. C1 (January 1995): C1–171—C1–175. http://dx.doi.org/10.1051/jp4:1995120.

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18

Bose, M. R. S. C., G. Thomas, R. Palaninathan, S. P. Damodaran, and P. Chellapandi. "Buckling investigations on a nuclear reactor inner vessel model." Experimental Mechanics 41, no. 2 (June 2001): 144–50. http://dx.doi.org/10.1007/bf02323190.

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19

Petrov, Yu I., B. F. Balunov, S. S. Davydovand, and E. E. Pakh. "Fusion reactor vacuum vessel cooldown." Plasma Devices and Operations 8, no. 3 (December 2000): 167–77. http://dx.doi.org/10.1080/10519990008228762.

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20

Kovbasenko, Yu, and Yevgen Bilodid. "Analysis of criticality of melt during severe accidents in reactor vessel." Nuclear and Radiation Safety, no. 2(78) (June 7, 2018): 3–10. http://dx.doi.org/10.32918/nrs.2018.2(78).01.

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The article investigates the possibility of a self-sustaining chain nuclear fission reaction during the development of a severe accident in the core at nuclear power plants with reactors WWER-1000 of Ukraine. Some models for calculating a criticality at different stages of the severe accident in the reactor VVER-1000 vessel were developed and calculations of multiplication properties of fuel containing masses were performed. The severe accident in the VVER-1000 core approximately divided into seven major stages: the intact reactor core, beginning of cladding damage (swelling), cladding melting and flowing down to the support grid, melting of constructional materials, homogenization of the materials at the bottom of the reactor vessel, stratification of corium at the bottom of the reactor vessel, the exit of the corium from the reactor shaft. It was shown that at the beginning of an accident, if fuel rods geometry is maintained, criticality might appear even if the emergency protection rods is triggered. With further development of the accident, the melt of fuel and structural materials will be deeply subcritical if water cannot penetrate into the pores or voids of the melt. In the case of the formation of pores or voids in the melt and the ingress of water into them, a recriticality may arise. A compensating measure is the addition of a boric acid solution to a cooling water with a certain concentration. According to the results of the computation analysis, a reactor core loaded with TVSA fuel (Russian production) requires a higher concentration of boric acid in water to compensate the multiplication properties of the fuel system in emergency situations compared to the core loaded with TVS-WR fuel (manufactured by Westinghouse), i.e. TVS-WR fuel is safer from the criticality point of view.

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21

Trampus, Péter. "Reactor Pressure Vessel Integrity in Light of the Evolution of Materials Science and Engineering." Materials Science Forum 473-474 (January 2005): 287–92. http://dx.doi.org/10.4028/www.scientific.net/msf.473-474.287.

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Structural integrity of the reactor pressure vessel of pressurized water reactors is one of the key safety issues in nuclear power operation. Integrity may be jeopardized during operational transients. The problem is compounded by radiation damage of the vessel structural materials. Structural integrity assessment as an interdisciplinary field is primarily based on materials science and fracture mechanics. The paper gives an overview on the service induced damage processes and associated changes of mechanical properties, the prediction of degradation and the assessment of the entire component against brittle fracture with a special focus on how the evolution of materials science and engineering has contributed to reactor vessel structural integrity assessment.

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22

Bakhracheva, Yulia S. "Influence of irradiation on the dynamics of changes in the mechanical properties of the nuclear reactor vessel." MATEC Web of Conferences 226 (2018): 01024. http://dx.doi.org/10.1051/matecconf/201822601024.

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Nuclear power plants are important generating units of the energy system worldwide. In the normal mode, nuclear power plants are absolutely safe, but emergency systems with radiation emissions have a devastating impact on the environment and public health. Despite the introduction of technologies and automatic monitoring systems, the threat of a potentially dangerous situation remains. The reactor vessel is the main object of activities to ensure the safety of nuclear power plants. One of the problems of ensuring the safety of nuclear power plant reactor vessels is the prediction of the level of crack resistance of reactor steels. The paper shows the possibility of estimating the neutron irradiation level on the nature of the temperature dependence of KIC. The prediction of the influence of radiation damage on the fracture toughness of the reactor steel can be obtained on the basis of the results of tests of small cylindrical samples with annular notches.

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23

Kim, J. C., Jae Boong Choi, Yoon Suk Chang, Young Jin Kim, Youn Won Park, and Young Hwan Choi. "Development of Web-Based Fatigue Life Evaluation System for Reactor Pressure Vessel." Solid State Phenomena 120 (February 2007): 25–30. http://dx.doi.org/10.4028/www.scientific.net/ssp.120.25.

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While the demand on electric power is consistently increasing, public concerns and regulations for the construction of new nuclear power plants are getting restrict, and also operating nuclear power plants are gradually ageing. For this reason, the interest on lifetime extension for operating nuclear power plants by applying lifetime management system is increasing. The 40-year design life concept was originally introduced on the basis of economic and safety considerations. In other words, it was not determined by technological evaluations. Also, the transient design data which were applied for fatigue damage evaluation were overly conservative in comparison with actual transient data. Therefore, the accumulation of fatigue damage may result in a big difference between the actual data and the design data. The lifetime of nuclear power plants is mostly dependent on the fatigue life of a reactor pressure vessel, and thus, the exact evaluation of fatigue life on a reactor pressure vessel is a crucial factor in determining the extension of operating life. The purpose of this paper is to introduce a real-time fatigue monitoring system for an operating reactor pressure vessel which can be used for the lifetime extension. In order to satisfy the objectives, a web-based transient acquisition system was developed, thereby, real-time thermal-hydraulic data were reserved for 18 operating reactor pressure vessels. A series of finite element analyses was carried out to obtain the stress data due to actual transient. The fatigue life evaluation has been performed based on the stress analysis results and, finally, a web-based fatigue life evaluation system was introduced by combining analysis results and on-line monitoring system. Comparison of the stress analysis results between operating transients and design transients showed a considerable amount of benefits in terms of fatigue life. Therefore, it is anticipated that the developed web-based system can be utilized as an efficient tool for fatigue life estimation of reactor pressure vessel.

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24

Isnaini, Muhammad Darwis, Elfrida Saragi, and Veronika Indriati Sri Wardani. "PREDICTION OF AP1000’S NUCLEAR REACTOR PRESSURE VESSEL TEMPERATURE DURING NORMAL OPERATION." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 24, no. 3 (November 9, 2022): 99. http://dx.doi.org/10.17146/tdm.2022.24.3.6684.

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Modeling of thermal-hydraulic calculations for the AP1000 core to predict the reactor pressure vessel (RPV) temperature has been carried out. The reactor’s primary coolant system transfers the heat produced in the reactor fuel during reactor operation to the steam generator. Part of the heat will also be transferred from the coolant to the reactor vessel and the pipe. This paper presents the calculation result of the RPV temperature prediction during AP1000 normal operation. Calculations were performed using COBRA-EN code for analyzing the core thermal hydraulics and using analytics for predicting the RPV temperature. These methods were carried out with the aim to predict the RPV temperature as well as at steady state nominal power conditions, at the function of flow, and at power fluctuation conditions. The calculation results at nominal power 3400 MWt (100% heat generated in fuel was assumed) and thermal design flow with 10% tube plugging (TDF2) of 48,443.7 ton/hr, for the minimum system pressure of 15.1 MPa, nominal system pressure of 15.513 MPa, and design system pressure of 17.133 MPa, show that the core outlet coolant temperature is 326.96°C, 327.01°C, and 327.22°C, and the RPV temperature is 303.65°C, 303.87°C, and 306.67°C, and the minimum departure from nucleate boiling ratio (MDNBR) is 3.21, 3.29, and 3.01, respectively. During reactor operation at a fixed nominal power of 3400 MWt, nominal system pressure, and under the condition of flow fluctuation, the maximum RPV temperature is shown to be 303.87°C.

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Kim, Hwan Yeol, Sevostian Bechta, Jerzy Foit, and Seong Wan Hong. "In-Vessel and Ex-Vessel Corium Stabilization in Light Water Reactor." Science and Technology of Nuclear Installations 2018 (October 2, 2018): 1–3. http://dx.doi.org/10.1155/2018/3918150.

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26

Fejt, Filip. "THERMAL-HYDRAULIC ANALYSIS OF IRT-4M IN REACTIVITY INSERTION ACCIDENT AT VR-1 REACTOR." Acta Polytechnica CTU Proceedings 4 (December 16, 2016): 22. http://dx.doi.org/10.14311/ap.2016.4.0022.

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The paper deals with thermal-hydraulic analysis during reactivity insertion accident, i.e. a step increase of nuclear system reactivity by 0.7 eff, at VR-1 Reactor. The reactor utilizes IRT-4M type of fuel assemblies, and even though these fuel assemblies are designed for an operation at the high-power research reactors, they might be also used for zero-power reactors. The thermal-hydraulic analyses must take into account several specific assumptions that are derived from VR-1 reactor specifications. The reactor does not require a forced water flow for a fuel cooling, the core is placed in an open vessel with atmospheric pressure, and amount of coolant water in the vessel is sufficient for providing the inlet water at room temperature for the whole event. Coolant circulation is expected to be formed only by natural convection.

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Weng, Yu, Haitao Wang, Haijun Wang, Jialun Liu, Jie Pan, and Zhian Deng. "The thermal-fluid-solid coupling effect in nuclear reactor vessel with direct vessel injection." Progress in Nuclear Energy 152 (October 2022): 104364. http://dx.doi.org/10.1016/j.pnucene.2022.104364.

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Gémes, György András. "Effect of Cladding on the Reactor Pressure Vessel Safety." Materials Science Forum 537-538 (February 2007): 363–70. http://dx.doi.org/10.4028/www.scientific.net/msf.537-538.363.

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It is common for the most reactor pressure vessel (RPV) of light-water reactors used today, that the RPV made of ferrite-perlitic steel is cladded with an austenitic layer, which well resists corrosion influence. The cladding is manufactured by welding, the technology (mode of application, number of layers) is different by each manufacturer. This is the layer which comes into contact with the working media first, and is of course in cohesive contact with the RPV material. The aim of this study is to demonstrate the new approach, used for assessing the safety of the reactors today. This approach plays a decisive role in the lifespan calculation of already operating nuclear power stations, including the extension of their service period.

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Pospíšil, Petr. "Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools." Technological Engineering 10, no. 2 (December 1, 2013): 6–10. http://dx.doi.org/10.2478/teen-2013-0012.

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Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

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Nguyen, Huu Tiep, Viet Ha Pham Nhu, and Minh Tuan Nguyen. "Calculation of neutron and gamma fluences on VVER reactor pressure vessel." Nuclear Science and Technology 6, no. 4 (December 30, 2016): 18–25. http://dx.doi.org/10.53747/jnst.v6i4.172.

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Embrittlement is one of the most important effects affecting reactor pressure vessel (RPV) aging. RPV is irradiated with neutrons and gammas, especially fast neutrons, which mainly lead to embrittlement of RPV during operation lifetime of nuclear reactors. Therefore, the radiation-induced embrittlement of the RPV should be carefully evaluated. In this paper, a preliminary calculation was performed using the MCNP5 code to identify the areas in the RPV of the VVER-1000/V320 reactor where the neutron and gamma fluxes are maximum. Also, the neutron and gamma fluence distributions on the RPV were investigated and evaluated along with their energy spectra. These calculations are the starting point for the evaluation of radiation damage to RPV of VVER reactors.

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31

Vuiart, Romain, Mariya Brovchenko, Julien Taforeau, and Eric Dumonteil. "A Detailed Analysis of the H.B. Robinson-2 Reactor Pressure Vessel Dosimetry Benchmark." Energies 15, no. 14 (July 12, 2022): 5098. http://dx.doi.org/10.3390/en15145098.

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The operation of many nuclear pressurized water reactors is being extended beyond their design lifetime limit. From the perspective of possible further lifetime extension, safety requirements are a priority. Therefore, the quantification of the neutron irradiation embrittlement of the reactor pressure vessel (RPV) is an important issue, as this is a guiding parameter that influences the reactor lifetime. In this context, the Institut de Radioprotection et de Sûreté Nucléaire developed a calculation scheme for the analysis of RPV aging under neutron irradiation, named VACS (vessel aging calculation scheme). VACS couples a deterministic approach (CASMO5 and SIMULATE5) to evaluate the full-core fission neutron source term and a Monte Carlo modeling (MCNP6) approach to model the neutron attenuation from the core to sites of interest (RPV, surveillance capsules, etc.). To ensure the reliability of aging predictions, this paper describes a detailed analysis of the neutron H.B. Robinson-2 reactor pressure vessel dosimetry benchmark. The results indicate that VACS shows satisfactory accuracy when the ENDF-B/VII.1 or JEFF-3.3 nuclear data libraries are used in the attenuation calculation. However, the use of ENDF-B/VIII.0 leads to significantly worse results.

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Dorf, Valery A., and Boris K. Pergamenchik. "Updating of dry shielding of nuclear power plant reactor vessel." Vestnik MGSU, no. 4 (April 2021): 506–12. http://dx.doi.org/10.22227/1997-0935.2021.4.506-512.

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Introduction. Dry shielding is a cylindrical structure made of serpentinite concrete in a metal casing with an inner diameter of 5.6 m, an outer diameter of 6.7 m, and a height of 5.3 m, which surrounds the VVER reactor vessel in the vicinity of the core. The purpose of serpentinite concrete, containing an increased amount of chemically bound water, is to soften the spectrum of the neutron flux outside the reactor, increasing the fraction of thermal neutrons in the spectrum, which is necessary for the operation of ionization chambers (IR) of the reactor control and protection system. Dry shielding also performs the functions of radiation and thermal protection, reducing the flux of radiation on ordinary concrete of biological protection. Before the installation of the dry shielding in the reactor shaft, heat treatment (drying) of concrete is carried out at temperatures up to 250 °С to remove unbound water in order to avoid radiolysis. Quality control of concreting and then heat treatment is carried out using a radioisotope device — a neutron moisture meter. These works are very lengthy and costly. Materials and methods. The design of the dry shielding casing was considered in order to perform additional perforation in order to avoid the formation of air pockets during concreting. The possibility of using modern plasticizing additives was considered in order to minimize the consumption of mixing water and, as a result, free water in the body of serpentinite concrete. Results. The possibility of exclusion of the stages of quality control of concreting and heat treatment in their traditional form is shown. Additional perforation of the metal casing, its internal diaphragms in problem areas, the use of a mixture of 20 cm slump or more allows you to completely eliminate the formation of internal voids. According to preliminary estimates, given the intensity of radiation in the NW for a modern reactor with a capacity of 1200 MW, the intensity of the release of hydrogen outside the shell due to radiolysis does not pose any danger. The concentration of hydrogen in the air surrounding the dry shielding is many orders less of magnitude than the dangerous 4 %. Conclusions. The cost of work on the construction of the SZ power unit of a nuclear power plant with a capacity of 1000–1200 MW can be reduced by 70–100 million rubles, the duration of work by 5 months.

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Bernsen, Sidney, Bryan Erler, Dana K. Morton, and Owen Hedden. "The Code Builders." Mechanical Engineering 136, no. 05 (May 1, 2014): 36–41. http://dx.doi.org/10.1115/1.2014-may-2.

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This article elaborates the evolution of code and standards for nuclear power plants. In the 1950s, need was felt for a revised set of design and fabrication rules to facilitate the development of safe, economically competitive water-cooled reactors contained in pressure vessels. These rules were codified in the first edition of the ASME Boiler and Pressure Vessel Code Section III, which was completed in 1963 and published in 1964. From the outset, both regulators and industry realized that the best way to develop many of the needed rules for the design, construction, and operation of nuclear facilities was the national standards consensus process. This process, followed by the American National Standards Institute and other recognized standards-issuing bodies such as ASME, brings together the expertise of individuals from government, industry, academia, and other stakeholders. In the years following the first publication of Section III, the coverage of the Code expanded to incorporate piping requirements, pressure-retaining components for pumps and valves, equipment and piping supports, reactor vessel internal structures, and other features of nuclear power plants.

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Nakata, Alexandre Ezzidi, Masanori Naitoh, and Chris Allison. "NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE." JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA 21, no. 3 (November 12, 2019): 119. http://dx.doi.org/10.17146/tdm.2019.21.3.5630.

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Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation

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Nguyen, Huu Tiep, Viet Ha Pham Nhu, and Minh Tuan Nguyen. "Investigation of DPA in the reactor pressure vessel of VVER-1000/V320." Nuclear Science and Technology 7, no. 4 (September 1, 2021): 16–25. http://dx.doi.org/10.53747/jnst.v7i4.93.

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The most important ageing effect on the reactor pressure vessel (RPV) is radiationembrittlement, which is mainly caused by fast neutrons during operation lifetime of nuclear reactors. The aim of this study was to investigate the DPA (displacement per atom) rate, an important parameter describing radiation damage to the RPV, and identify the position of the maximum DPA rate in the RPV of the VVER-1000/V320 reactor using the Monte Carlo code MCNP5. To reduce statistical errors in the MCNP5 simulation, the weight window technique was applied to non-repeated structures outside the reactor core. The results showed the distribution of the DPA rate in the RPV and the maximum DPA rate was found to be at the first millimeters of the RPV. Consequently, these calculations could be useful for assessment of radiation damage to the RPV of VVER reactors.

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Park, Sang-Hyun, Kwang-Hyun Bang, and Jong-Rae Cho. "Structural Integrity Evaluation of a Reactor Cavity during a Steam Explosion for External Reactor Vessel Cooling." Energies 14, no. 12 (June 17, 2021): 3605. http://dx.doi.org/10.3390/en14123605.

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Nuclear power is a major source of electricity in the international community. However, a significant problem with nuclear power is that, if a severe nuclear accident occurs, radiation may leak and cause great damage. As such, research on nuclear safety has become increasingly popular worldwide. In this paper, the structural integrity of a reactor cavity during a steam explosion—one kind of the aforementioned severe nuclear accidents—was evaluated. Steam explosions are primarily caused by fuel–coolant interactions (FCI), and result from issues in the cooling system that discharges the melt from the reactor core to the outside. A steam explosion can damage the nuclear power plant, and radiation leakage, the greatest concern, may occur. In the Chernobyl or Fukushima Daiichi accidents, significant radiation leakages resulted in damages extending beyond the country of origin. In this paper, a steam explosion was simulated using values given by the transient analysis code for explosive reactions (TRACER-II)—the only steam explosion code in Korea. The walls of the reactor cavity were modeled after the APR-1400 currently operating in Korea. The integrity of the concrete, rebars, and liner plate in the reactor cavity during a steam explosion was evaluated in terms of stress and ductile failure strain limits.

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Shiga, Tomoya, Yudai Tanaka, and Tetsuaki Takada. "Process of Air Ingress during a Depressurization Accident of GTHTR300." Science and Technology of Nuclear Installations 2018 (September 2, 2018): 1–15. http://dx.doi.org/10.1155/2018/6378504.

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A depressurization accident is the design-basis accidents of a gas turbine high temperature reactor, GTHTR300, which is JAEA’s design and one of the Very-High-Temperature Reactors (VHTR). When a primary pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. Therefore, it is important to know a mixing process of different kinds of gases in the stable and unstable density stratified fluid layer. In order to predict or analyze the air ingress phenomena during the depressurization accident, we have conducted an experiment to obtain the mixing process of two component gases and the characteristics of natural circulation. The experimental apparatus consists of a storage tank and a reverse U-shaped vertical rectangular passage. One side wall of the high temperature side vertical passage is heated and the other side wall is cooled. The other experimental apparatus consists of a cylindrical double coaxial vessel and a horizontal double coaxial pipe. The outside of the double coaxial vessel is cooled and the inside is heated. The results obtained in this study are as follows. When the primary pipe is connected at the bottom of the reactor pressure vessel, onset time of natural circulation of air is affected by not only molecular diffusion but also localized natural convection. When the wall temperature difference is large, onset time of natural circulation of air is strongly affected by natural convection rather than molecular diffusion. When the primary pipe is connected at the side of the reactor pressure vessel, air will enter the bottom space in the reactor pressure vessel by counter-current flow at the coaxial double pipe break part immediately. Afterward, air will enter the reactor core by localized natural convection and molecular diffusion.

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Langston, Lee S. "Pebbles Making Waves." Mechanical Engineering 130, no. 02 (February 1, 2008): 34–38. http://dx.doi.org/10.1115/1.2008-feb-3.

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This paper describes various high-level nuclear researches including nuclear-fuelled pebbles that are being conducted across South Africa. The pebbles are ingenious industrial products, designed to passively limit the amount of heat unleashed by the nuclear fission reactions that drive the reactor. The spheres that give the pebble bed reactor its name enclose fissionable uranium inside layers that serve various roles, such as moderating fission, containing pressure, and accommodating deformation of the core. Nuclear-fuelled pebbles are introduced at the top of the reactor vessel and slowly wend their way down through the annular-packed bed under the action of gravity to the bottom of the reactor vessel. In a towering building at the headquarters of Nesca in Pelindaba, South Africa, reactor components are being tested for their ability to work with high-pressure helium. Those parts will go in the pebble bed modular reactor power plant to be constructed at Koeburg, near Cape Town. The plan of the pebble bed reactor power plant will use the helium coolant to run the turbine directly rather than heat a secondary fluid, as in a water reactor.

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Yahr, G. T. "Fatigue Design Curves for 6061-T6 Aluminum." Journal of Pressure Vessel Technology 119, no. 2 (May 1, 1997): 211–15. http://dx.doi.org/10.1115/1.2842286.

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A request has been made to the ASME Boiler and Pressure Vessel Committee that 6061-T6 aluminum be approved for use in the construction of Class 1 welded nuclear vessels so it can be used for the pressure vessel of the Advanced Neutron Source research reactor. Fatigue design curves with and without mean stress effects have been proposed. A knock-down factor of 2 is applied to the design curve for evaluation of welds. The basis of the curves is explained. The fatigue design curves are compared to fatigue data from base metal and weldments.

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Park, Rae-Joon, Jae Ryong Lee, Kwang Soon Ha, and Hwan Yeol Kim. "Evaluation of in-vessel corium retention through external reactor vessel cooling for small integral reactor." Nuclear Engineering and Design 262 (September 2013): 571–78. http://dx.doi.org/10.1016/j.nucengdes.2013.06.003.

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Ara, K., N. Ebine, and N. Nakajima. "A New Method of Nondestructive Measurement for Assessment of Material Degradation of Aged Reactor Pressure Vessels." Journal of Pressure Vessel Technology 118, no. 4 (November 1, 1996): 447–53. http://dx.doi.org/10.1115/1.2842212.

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A method, MIM (magnetic interrogation method), is proposed for nondestructive measurement of radiation damage of nuclear reactor pressure vessels. The method relies on good correlation between the levels of radiation-induced hardness change and magnetic coercivity change in pressure vessel steel. A part of the pressure vessel to be inspected is magnetized with two-pole magnetic yokes through the overlay clad of nonmagnetic stainless steel, and magnetic field distributions on the surface of overlay clad are measured in the vicinity of the poles of magnetic yokes. Then, the coercivity distribution in the direction of thickness in the pressure vessel steel is inversely estimated from the measured magnetic field distribution patterns with aid of static magnetic field analysis. The level of radiation damage such as hardness change is assessed with relation to the estimated coercivity distributions.

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42

Rodríguez-Prieto, Alvaro, Manuel Callejas, Ernesto Primera, Guglielmo Lomonaco, and Ana María Camacho. "Multicriteria Analytical Model for Mechanical Integrity Prognostics of Reactor Pressure Vessels Manufactured from Forged and Rolled Steels." Mathematics 10, no. 10 (May 23, 2022): 1779. http://dx.doi.org/10.3390/math10101779.

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The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-service of reactor pressure vessels. The results obtained by the application of the methodology are coherent with several historical experimental works.

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Yang, J., M. B. Dizon, F. B. Cheung, J. L. Rempe, K. Y. Suh, and S. B. Kim. "CHF enhancement by vessel coating for external reactor vessel cooling." Nuclear Engineering and Design 236, no. 10 (May 2006): 1089–98. http://dx.doi.org/10.1016/j.nucengdes.2005.11.008.

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Wei, Hui Ming, Xuan Zhang, and Tai Li Liu. "AP1000 Nuclear Reactor and Primary Loop Modeling Based on Relap5-3D and 3Keymaster Simulation Platform." Applied Mechanics and Materials 214 (November 2012): 510–14. http://dx.doi.org/10.4028/www.scientific.net/amm.214.510.

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In this paper, the advanced 3Keymaster simulation platform and RELAP5 simulation code are introduced. At the same time, the AP1000 nuclear reactor and primary loop modeling method is discussed in detail based on Remlap 5 and 3Keymaster simulation platform. The typical Relap5 nasalization cells of AP1000 nuclear reactor and primary loop system are showed in Fig.1. The Fig.2 shows the AP1000 reactor coolant system model. Fig.3 shows the AP1000 reactor core and pressure vessel model. Fig.4 shows the AP1000 nuclear reactor steam generator model.

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45

Van Der Sluys, W. A., and R. H. Emanuelson. "Cyclic Crack Growth Behavior of Reactor Pressure Vessel Steels in Light Water Reactor Environments." Journal of Engineering Materials and Technology 108, no. 1 (January 1, 1986): 26–30. http://dx.doi.org/10.1115/1.3225836.

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During normal operation light water reactor (LWR) pressure vessels are subjected to a variety of transients resulting in time varying stresses. Consequently, fatigue and environmentally assisted fatigue are growth mechanisms relevant to flaws in these pressure vessels. In order to provide a better understanding of the resistance of nuclear pressure vessel steels to flaw growth process, a series of fracture mechanics experiments were conducted to generate data on the rate of cyclic crack growth in SA508-2 and SA533B-1 steels in simulated 550°F Boiling Water Reactor (BWR) and 550°F Pressurized Water Reactor (PWR) environments. Areas investigated over the course of the test program included the effects of loading frequency and R ratio (Kmin/Kmax) on crack growth rate as a function of the stress intensity factor (ΔK) range. In addition, the effect of sulfur content of the test material on the cyclic crack growth rate was studied. Cyclic crack growth rates were found to be controlled by ΔK, R ratio, and loading frequency. The sulfur impurity content of the reactor pressure vessel steels studied had a significant effect on the cyclic crack growth rates. The Higher growth rates were always associated with materials of higher sulfur content. For a given level of sulfur, growth rates were higher in a 550°F simulated BWR environment than in a 550°F simulated PWR environment. In both environments cyclic crack growth rates were a strong function of the loading frequency. Further, the loading frequency at which the highest cyclic crack growth rate was observed was found to be a function of the applied ΔK level. In most cases, all cyclic crack growth rates were on or under the ASME Section XI high R water reference flaw growth line and above the Section XI air reference flaw growth line, supporting the position of these lines on the growth rate–ΔK level graph.

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46

Paulech, Juraj, Vladimír Kutiš, Gabriel Gálik, Jakub Jakubec, and Tibor Sedlár. "Thermo-Hydraulic Behaviour of Coolant in Nuclear Reactor VVER-440 Under Refuelling Conditions." Strojnícky casopis – Journal of Mechanical Engineering 67, no. 1 (April 1, 2017): 87–92. http://dx.doi.org/10.1515/scjme-2017-0009.

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Abstract The paper presents the numerical simulation of thermo-hydraulic behaviour of coolant in the VVER- 440 nuclear reactor under standard outage conditions. Heating-up and flow of coolant between the reactor pressure vessel and spent fuel storage pool are discussed.

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Liao, Jun, and Dan Utley. "Study on Reactor Vessel Air Cooling for Westinghouse Lead Fast Reactor." Nuclear Technology 206, no. 2 (April 29, 2019): 191–205. http://dx.doi.org/10.1080/00295450.2019.1599614.

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Popov, V., V. Mileikovskyi, and О. 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 cyclic damage." Ventilation, Illumination and Heat Gas Supply 39 (December 6, 2021): 6–28. http://dx.doi.org/10.32347/2409-2606.2021.39.6-28.

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Ukraine remains a country dependent on nuclear energy for both heat supply and satellite supply. In cities close to nuclear power plants, electric heating from the electricity they produce is appropriate. On the other hand, Ukraine is the only country in the world where the worst accident at a nuclear power plant occurred at the Chernobyl NPP Unit 4 on the night of April 26, 1986. Another characteristic feature of Ukraine's nuclear energy is the significant number of power units with exhausted project resources - the so-called "old" power units. Their wear is associated with the influence of heat and mass transfer processes, which lead to periodic thermal deformation of the elements, which causes cyclic damage. An example of expert rapid assessment of the residual life of a specific reactor vessel WWER-1000 is given taking into account the combined action of non-stationary heat and mass transfer and mechanical processes. A detailed express calculation of cyclic (tired) damage to the metal of the WWER-1000 reactor vessel due to the dangerous emergency mode has been performed. The WWER-1000 reactor is the last most widespread and most powerful water-water nuclear reactor of the former USSR, operated on 13 of the 15 operating power units of Ukrainian nuclear power plants (NPPs) - Zaporizhia, Rivne, Khmelnytsky and South Ukraine (SUNPP). The reactor is the most important indispensable element of a nuclear power plant, which determines its safety and resource. Given the design resource and the dates of commissioning of Ukrainian NPP units, the issue of rapid expert assessment of the technical condition of WWER-1000 hulls is quite relevant. Of course, modern engineering has the full range of tools needed to perform such estimates, from powerful computers to advanced computing software. But the known and inevitable costs of modern engineering - complex and time-consuming modeling and calculations. Experience has shown that a certain "reasonable" combination of the use of relatively simplified "light" calculation methods allows us to assess the safety, strength and service life of WWER-1000 reactors very quickly and with the necessary accuracy acceptable for expert opinion. The publication considers as an example the real emergency situation of October 22, 1985 and the WWER-1000 reactor of SUNPP Unit 1. Since this emergency situation led to rapid cooling of this reactor, the calculation of the effect of cyclic damage of reactor steel on the strength and reliability of the reactor vessel is shown. This example and the consideration of the real emergency situation demonstrates the effectiveness and acceptability of the use of estimated expert rapid assessments to accurately determine the reliability and safety of such critical elements of nuclear power plants as nuclear reactor buildings.

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Sehgal, B. R., A. Karbojian, A. Giri, O. Kymäläinen, J. M. Bonnet, K. Ikkonen, R. Sairanen, et al. "Assessment of reactor vessel integrity (ARVI)." Nuclear Engineering and Design 235, no. 2-4 (February 2005): 213–32. http://dx.doi.org/10.1016/j.nucengdes.2004.08.055.

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Sehgal, B. R., A. Theerthan, A. Giri, A. Karbojian, H. G. Willschütz, O. Kymäläinen, S. Vandroux, et al. "Assessment of reactor vessel integrity (ARVI)." Nuclear Engineering and Design 221, no. 1-3 (April 2003): 23–53. http://dx.doi.org/10.1016/s0029-5493(02)00343-6.

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Journal articles on the topic 'Nuclear reactor vessel'

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