Results 1 - 10 of 5894
Results 1 - 10 of 5894. Search took: 0.025 seconds
|Sort by: date | relevance|
[en] In Indian Pressurised Heavy Water Reactors (IPHWRs) calandria tubes are rolled with end shield tube sheet at either ends with the help of a sandwich type joints. These calandria tubes are generally kept unchanged during en-masse replacement of pressure tubes and end-fittings; probably due to unavailability of technology for replacement of calandria tube. But in recent past, one irradiated calandria tube has been replaced successfully from one of the 220 MWe IPHWRs, with a new one, with the help of Calandria Tube Rolled Joint Detachment (CTRJD) system, developed by Reactor Engineering Division of Bhabha Atomic Research Centre (BARC). This paper gives brief description of CTRJD system, methodology of calandria tube rolled joint detachment, shop floor trials of the technique for optimisation of operating parameters, qualification trials at full length mock up facility and deployment at reactor site. (author)
[en] After a loss-of-coolant accident (LOCA) in a Pressurized Water Reactor (PWR), the temperature of the fuel elements cladding increases dramatically due to the heat produced by the fission products decay, which is not adequately removed by the vapor contained in the core. In order to avoid this sharp rise in temperature and consequent melting of the core, the Emergency Core Cooling System is activated. This system initially injects borated water from accumulator tanks of the reactor through the inlet pipe (cold leg) and the outlet pipe (hot leg), or through the cold leg only, depending on the plant manufacturer. Some manufacturers add to this, direct injection into the upper plenum of the reactor. The penetration of water into the reactor core is a complex thermofluidodynamic process because it involves the mixing of water with the vapor contained in the reactor, added to that generated in the contact of the water with the still hot surfaces in various geometries. In some critical locations, the vapor flowing in the opposite direction of the water can control the penetration of this into the core. This phenomenon is known as Countercurrent Flow Limitation (CCFL) or Flooding, and it are characterized by the control that a gas exerts in the liquid flow in the opposite direction. This work presents a proposal to use a CFD to simulate the CCFL phenomenon. Numerical computing can provide important information and data that is difficult or expensive to measure or test experimentally. Given the importance of computational science today, it can be considered a third and independent branch of science on an equal footing with the theoretical and experimental sciences. (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] 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] As part of a research and development project funded by the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU), GRS has been developed a generic analysis simulator for research reactors of an open-pool type with a compact core. Aim of the project was the development of methods to analyse research reactors. Focus was on the methodological performance of deterministic safety analyses.This also included the approach to develop the input deck based on to the GRS simulation code ATHLET. Within the mode development of the reference plant the data were only available to a limited extent which required in case of need best guess assumptions. The development of the thermal-hydraulic model comprises the primary cooling system including the reactor core, pipes of the primary coolant system, check valves, pumps and heat exchanger. The secondary system is modelled in a simplified manner using boundary conditions (with fill and time-dependent elements) due to unknown data. The emergency cooling system is modelled with three loops and pumps as well as the reactor basin. Further, the natural circulation flaps between the reactor basin and the primary coolant system are modelled as well as a simple replication of the reactor building is given. The model of the I&C systems comprises basic functionalities of relevant operational and safety-oriented systems. This includes a simplified logic for the shut-down of the main coolant pumps as well as the controlling of the emergency cooling system. Furthermore, the characteristic of the opening of the natural circulation valves is modelled by a logical controller. The simulation of the reactor scram is based on a simplified model. The developed plant model has been tested by a stationary and a transient simulation. The simulated plant behavior is plausible, and the obtained results are in good agreement to the reference values. Finally, two events with a partial blockage of a core channel has been performed and described.
[de]Im Rahmen eines vom Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit (BMU) geförderten Eigenforschungs- und Entwicklungsvorhabens wurde ein generischer Analysesimulator für einen Forschungsreaktor vom sogenannten „Open-Pool“-Typ mit einem Kompaktkern erstellt. Ziel dieses Projekts war die Entwicklung von Methoden zur Analyse von Forschungsreaktoren. Hierbei stand die methodische Durchführung deterministischer Sicherheitsanalysen im Vordergrund. Dies schließt insbesondere den Vorgang der Entwicklung eines Eingabedatensatzes ein. In der Modellerstellung wurdendie begrenzt vorliegenden Daten zur Referenzanlage im Bedarfsfall durch Annahmen ergänzt. Die thermohydraulische Modellerstellung umfasst das Primärkühlsystem mit dem Reaktorkern, Rohrleitungen des Primärkühlsystems, Rückschlagklappen, Pumpen und Wär-metauscher. Das Sekundärkühlsystem ist aufgrund fehlender Daten vereinfacht da-gestellt und mit Randbedingungen (Fill und Time-Dependent Volume) versehen. Das Not- und Nachkühlsystem ist mit drei Strängen und Pumpen sowie dem Reaktorbecken modelliert. Weiterhin sind die Naturumlaufklappen als Verbindung zwischen Reaktorbecken und Primärkreislauf modelliert. Ebenso wird das Reaktorgebäude vereinfacht dargestellt. Die leittechnische Modellierung umfasst Grundfunktionalitäten relevanter betrieblicher und sicherheitstechnischer Systeme. Dies schließt eine vereinfachte Logik zum Abschalten der Hauptkühlmittelpumpen sowie eine vereinfachte Logik zur Steuerung der Not- und Nachkühlpumpen ein. Weiterhin ist das Öffnungsverhalten der Naturumlaufventile als Logikbaustein nachgebildet. Zur Simulation der Reaktorschnellabschaltung wird ein vereinfachtes Reaktorschutzsystem modelliert. Die Modellierung wird mit einer stationären und transienten Simulation überprüft. Das Anlagenmodell gibt Referenzwerte gut wieder und verhält sich plausibel. Abschließend sind zwei Störfallsimulationen mit einem teilweisen blockierten Kernkanal durchgeführt worden.
[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 the light of Fukushima accident scenario, there has been a great deal of concern regarding the problem of hydrogen production and combustion in Light Water Reactor (LWR) severe accidents. Hydrogen combustion in Nuclear Power Plant (NPP) may threaten the integrity of the containment boundary represents the final barrier to release of radioactivity to the environment. In this paper the hydrogen distribution analysis of Chasma Nuclear Power Plant Unit-1 (C-l) containment has been performed for the retrofitting of Passive Auto-catalytic Recombiners (PARs) in light of Fukushima Response Action plans (FRAP) to mitigate hydrogen deflagration or detonation risk in severe accidents. The hydrogen distribution analysis with PARs and without PARs was to check the hydrogen management capability to keep hydrogen concentration within permissible regulatory limits. Hydrogen distribution analysis for large/small break Loss of Coolant Accident (LOCA) and total loss of feed-water coincident with the failure of Emergency Core Cooling system (ECCS) are performed using MELCOR code. The 300 MWe Pressurized Water Reactor (PWR) with large dry containment is divided into 53 control volumes to determine the number and location/elevation of PARs. The results show that a certain number of recombiners could remove effectively hydrogen to protect the containment integrity against hydrogen deflagration or detonation. If adequate preventive and mitigative strategies to cope with severe accidents are not implemented by using PARs and accident management procedures, then such sequences may lead to core melt and failure of the ultimate barrier, i.e. the reactor containment and release of radioactivity to the environment. (author)
[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] Fuel heat-up, fuel degradation in an accident and the resultant fuel failure with release of the fission product (FP) into the primary system during Design Basis Accident (DBA) and Design Extension Conditions (DEC) are the key aspects to demonstrate the safety of the NPPs. Post Fukushima more emphasis is also laid on the development of the Accident Tolerant Fuel Designs (ATFD) to avoid fuel degradation and hydrogen generation. Terms like practically eliminated and ATF, need to be substantiated with physical and analytical evidence. Along with ATF development efforts should also be dedicated in prevention of loss of heat removal and/or quick restoration and lining up of the emergency coolant inventories with in the capabilities/survivability of the ATF. The aspects related to DBA and DEC fuel/core modelling are evolving specially the later one. The expectations, development, is also discussed here along with the comparison of ATF with the existing fuels. The development of ATF may be an iterative process shuffling from nuclear requirements to materials and safety performance while testing out of pile and in-pile. ATF aspect w.r.t reactivity loads is also elaborated for PHWRs. The developments on RIA and high burn-up fuel are also summarised. With improved ATF fuel performance improved MHT design and configuration are also envisaged. It is possible to configure MHT, ECCS and passive safety system in such a manner that the possibility of prolonged loss of cooling is reduced. (author)
[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.