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[en] The presented in this paper results of neutron-physical, thermophysical, and technological studies have shown that it is possible to provide the required parameters of a high-temperature reactor installation with a sodium coolant for hydrogen production and other innovative applications based on one of thermochemical cycles or high temperature electrolysis with a high coefficient of thermal use of electricity
[ru]Представленные в работе результаты нейтронно-физических, теплофизических и технологических исследований показали, что имеется принципиальная возможность обеспечить требуемые параметры высокотемпературной реакторной установки с реактором с натриевым теплоносителем для производства водорода и других инновационных приложений, на основе одного из термохимических циклов или высокотемпературного электролиза с высоким коэффициентом теплового использования электроэнергии
[en] During the assessment of the static strength of the flange connections elements Dn2130 and Dn2080 of the emergency cooling heat exchangers 08.8111.335 SB (TOAR), it was found that there is an excess of the allowable stress values. These calculations of static strength per-formed using the finite element method (FEM). The analysis of the static strength of the flange joints was performed taking into account the design values of the tightening of the studs, equal to 22,527 kgf and 8,836 kgf, accordingly. At the same time, one of the main purposes of heat exchangers TOAR nuclear installation (NI) WWER-1000 is the work until accidents. The analysis of accidents of NI WWER-1000 showed that the largest values of change of parameters of environments in heat exchangers of TOAR correspond to accident “LOCA: Bilateral rupture of MCT”. Based on this, we considered the thermal stress state of heat exchangers for this accident. To determine the thermal stress state of the TOAR heat exchanger elements, during accidents of the nuclear installation, strength calculations were performed in the non-stationary formulation of the problem. One of the boundary conditions for these strength calculations is the distribution of temperatures along the thickness and length of the walls of the elements of the heat exchanger, which changes over time. Numerical thermohydraulic calculations were performed to determine these boundary conditions In the article for the first time the results of calculations of thermal stress state of separate elements of heat exchangers TOAR, for work of heat exchangers during accidents of nuclear installation are received. It is established that the elements of the flange connection Dn2130 are one of the most critical elements of TOAR heat exchangers. To determine the thermal stress state of the heat exchanger elements, analytical thermal calculations, numerical thermohydraulic and strength calculations were performed using the FEM method. As a result of the analysis of the performed strength calculations, it was concluded that it is necessary to reduce the tightening value of the flanges of the flange connection Dn2130 to 14600 kgf. (author)
[en] The design strength of 61409 RR1 emergency cooling of the heat exchanger 08.8111.335 SB, as the main design and factory document governing the safe operation of the heat exchanger during its operation in such modes as normal operating conditions, hydraulic tests and seismic loads under time of normal operating conditions is considered and analyzed in the article. The purpose of the work is to analyze the document 61409 RR1 for compliance with current standards of Ukraine in nuclear energy. It is shown that the design strength calculation 61409 RR1 doesn't comply with the requirements of current regulatory documents. The document does not present the results of the calculation of static and cyclic strength for the elements of flange joints and studs in particular. However, the results of the calculations of the studs, given in the section “ Structural calculation” demonstrate the excess of the allowable values of stresses in the group of membrane stresses. Since 2016, a new normative document NP 306.2.208- 2016 has been in force in Ukraine, which replaced the norms of PNAE G-5-006-87. The new normative document states that one of the combinations of loads, when considering seismic effects, is violation of normal operating conditions and maximum considered earthquake. Therefore, document 61409 PP1 can not be used as a technical document regulating the safe operation of the emergency cooling heat exchanger 08.8111.335 SB during seismic impacts in works related to the justification of safe operation of equipment of existing NPPs of Ukraine. Based on the above, it is recommended to perform additional calculations on the strength of the emergency heat exchanger 08.8111.335 SB, which will also take into account the calculations of the elements of flange connections, as well as a combination of violations of normal loads and the maximum predicted earthquake, and generally meet current regulations of Ukraine in nuclear energy. (author)
[en] ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.
[en] In this paper, the design basis accident (DBA) of a BWR/5 evolving to severe accident, by disabling all ECCSs, is the scenario chosen to study hydrogen generation in core, and then to determine the hydrogen distribution in the drywell. Alternative scenarios are simulated by allowing the recovery of the high pressure cooling system (HPCS) at different times during this scenario. The code RELAP/SCDAPSIM was used to simulate the severe accident scenarios, and the code GASFLOW was used to compute the 3D hydrogen distribution in a drywell of a Mark II primary containment. Results of hydrogen and fission product generation in core, as well as state of core degradation are presented. Then, a comparison of results of hydrogen concentration in drywell between the base case and the cases where coolant injection is recovers is presented. (author)
[en] To evaluate the fracture resistance of high-burnup advanced fuel cladding tubes during the long-term core cooling period following loss-of-coolant accidents (LOCAs), laboratory-scale four-point-bend tests were performed using the following advanced fuel cladding tubes with burnups of 73-84 GWd/t: low-tin ZIRLOTM, M5®, and Zircaloy-2 (LK3). Three four-point-bend tests were performed on the high-burnup advanced fuel cladding tube specimens subjected to the integral thermal shock tests which simulated LOCA conditions (ballooning and rupture, oxidation in high-temperature steam, and quench). During the four-point-bend tests, all the specimens that were oxidized at 1474 K to 9.9%-21.5% equivalent cladding reacted exhibited brittle fractures. The maximum bending moments were comparable to those of the conventional Zircaloy cladding tube specimens. Furthermore, the effects of oxidation and hydriding on the maximum bending moment were comparable between the high-burnup advanced fuel cladding tube specimens and the unirradiated Zircaloy-4 cladding tube specimens. Therefore, it can be concluded that the post-LOCA fracture resistance of fuel cladding tubes is not significantly reduced by extending the burnup to 84 GWd/t and using the advanced fuel cladding tubes, though it may slightly decrease with increasing initial hydrogen concentration in a relatively lower ECR range (< 15%), as observed for the unirradiated Zircaloy-4 cladding tubes. (author)
[en] Hitachi-GE Nuclear Energy Ltd. set forth a nuclear power vision to accomplish the following purposes: (1) reduction in the risk of initial investment, (2) securement of long-term stable power sources, and (3) reduction in the harmfulness of radioactive waste. To realize these purposes, it is developing three types of advanced reactors: (1) small light water reactor BWRX-300, (2) light water cooling fast reactor RBWR, and (3) small liquid metal cooling fast reactor PRISM. BWRX-300 achieves both safety and economic efficiency through radical simplification, RBWR realizes fast reactor based on well-experienced light water cooling technology, and PRISM achieves high intrinsic safety and economic efficiency by adopting innovative technology. In the future, this company plans to carry out technological development to provide solutions to global energy problems, aiming at the practical use of these three reactor types at an early stage. (A.O.)
[en] Physical and chemical effects of containment debris on the performance of emergency coolant recirculation are investigated to get insight on the cost-effective plant modifications to resolve USNRC's Generic Safety lssue-191. The effects of debris sources on the sump screen performance are evaluated through the head loss calculation using NUREG/CR-6224 correlation. The amount of three predominant types of precipitates, i.e., sodium aluminum silicate (NaAlSi_3O_8), aluminum oxyhydroxide (AIOOH), calcium phosphate (Ca_3(PO_4)_2) after 30 days of ECCS mission time are evaluated under various environmental conditions using WCAP-16530-NP chemical models. The debris interceptor is considered as a viable design option to reduce particulate debris such as unqualified coatings. The key parameters of each effect are deduced and recommendations for reducing their adverse effects are made through the present analysis: (a) The amount of unqualified coating debris is a major source of particulate debris and has a great adverse effect on the sump screen head loss by reducing porosity in the fibrous insulation, (b) The Cal-Sil insulation reacts with TSP buffer and significantly increases the generation of a gum-like chemical precipitant, (c) Spray time increases the chemical byproducts but the effect is smaller than that of buffer agent type and unqualified coating, (d) The debris interceptor, when verified, may play a vital role reducing head loss generated by coatings and fibrous debris mix.
[en] For the reactor pressure vessel (RPV) integrity assessment, deterministic fracture mechanics (DFM) is applied considering factors such as neutron irradiation embrittlement, postulated crack and thermals transients etc. On the other hand, probabilistic fracture mechanics (PFM) can evaluate the failure probability considering the uncertainties of several kinds of factors affecting the RPV integrity. The purpose of present study is to obtain insight concerning the further improvement of RPV structural integrity from the probabilistic approach consideration through careful comparisons with DFM approach. It was found that the transients which give rise to the re-pressurization have a possibility to deteriorate the deterministic – probabilistic correlation. (author)
[en] When a heat transfer tube of the steam generator of a pressurized water reactor fails, the primary cooling water leaks quickly into the secondary system. Moreover, if this leakage is large, the nuclear reactor emergency core cooling system (ECCS) may be activated. In Japan, to prevent such situation to take place, periodic inspections are performed in order to check whether heat transfer tubes are cracked. Eddy Current Testing (ECT) is a type of non-destructive inspection method used to detect cracks in a conductive material. ECT can estimate the shape of a crack by inverse problem analysis, but it is computationally expensive. Therefore, in this study, we aimed to develop a method to estimate crack depth by Convolutional Neural Network (CNN). The method was shown to be less computationally expensive during estimation and was robust against lift-off fluctuation during measurements. (author)