Results 1 - 10 of 66710
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[en] The effects of adding warm plasmas on the kinetic DCLC mode in high β loss cone plasmas are investigated in detail. It is found that when the fluid DCLC mode is stabilized by a small amount of warm plasma, the kinetic excitation still remains due to two different mechanisms, namely, (1) magnetic drift resonance dissipation excites the negative energy wave; (2) a new type of positive energy wave can become unstable as the resonance condition is met. Comparing with fluid approximation theory, more warm plasmas are needed to suppress the kinetic DCLC instabilities
[en] The role of the functional elements of the compact torus system in suppressing energy and particle losses from the plasma is demonstrated. Experiments are carried out using the TOR and BN facilities with B=7-11kG, n=(2-4)x1015cm-3 using compensated diamagnetic loops, radially inserted magnetic probes, neutron and X-ray detectors and an He-Ne laser interferometer (∫ndl). It is established that power losses of magnetic flux and energy develop under conditions of spontaneous shaping of the closed structure or when the control elements are partially utilized. A marked improvement in all the characteristics of the compact torus is observed when programmed operation of the control elements is optimized (in accordance with the compact torus concept developed at the Kurchatov Institute). An appreciable increase in energy inside the separatrix is observed when the start of the longitudinal wave is delayed; the evolution of the geometrical dimensions and parameters of the plasma during compression satisfactorily agrees with two-dimensional MHD calculations assuming zero losses. When the external field is reduced, an increase in the radius and length of the compact torus is observed, which to within measurement error obeys the adiabatic law. An ion temperature of T approx.= 1.4keV is achieved, and the collisionless nature of the heating (lambdasub(ii)/L approx.= 50) is confirmed. It is established that an unfavourable torus structure which initiates losses may be self-sustaining. (author)
[en] Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accidents such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This paper proposes a new helium-cooled, tritium breeding blanket concept which uses a metallic structure, and which performs significantly better during such accidents than related designs. The proposed blanket uses modified, reduced-activation HT-9 steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. This concept uses novel features such as: (1) a open-quotes beryllium-jointclose quotes design which allows beryllium to be used to conduct heat away from the first wall, while accomodating swelling of the beryllium, and (2) a shield cooled by naturally circulating water. These features help the blanket passively withstand a worst-case undercooling accident scenario
[en] A Loss of Offsite Power (LOOP) transient scenario is based on a complete loss of non-emergency AC power that results in the loss of all power to the plant auxiliaries, i.e., the Reactor Coolant Pumps (RCPs), condensate pumps, etc. An actual LOOP event would cause a loss of all feedwater, a loss of forced Reactor Coolant System (RCS) flow and a reactor trip within less than 2 seconds as a result of either loss of power to the rod cluster assembly gripper coils or any RCS flow trips. For safety analysis purposes the LOOP event is conservatively modelled as a Loss of Normal Feedwater (LONF) transient with a subsequent loss of offsite power as a result of a reactor trip. The reactor trip followed by RCP trip are delayed until a low-low Steam Generator (SG) level signal is reached. This is a more conservative scenario than the LOOP event because the least amount of SG secondary side water mass available for heat removal and the increased amount of the stored energy in the primary circuit at the time of the loss of RCS flow result. The standard LOOP safety analysis is aimed to demonstrate the natural circulation capability of the RCS to remove residual and decay heat from the core aided by Auxiliary Feedwater in the secondary system. In addition to this goal the presented work is aimed to resolve the potential safety issue resulting from the influence of the Chemical and Volume Control System (CVCS) operation during LOOP event for NPP Krsko. The potential safety concern for the LOOP analysis is that the loss of instrument air system may occur thus leading to the CVCS charging and letdown flow imbalance. A net RCS inventory addition may result with water solid pressurizer condition. Water discharge through the pressurizer relief and safety valves could lead to overpressurization of the Pressurizer Relief Tank (PRT) and rupture of the PRT rupture disks. Additional concern is that pressurizer relief and safety valves may fail to properly reseat when exposed to water relief causing the American Nuclear Society (ANS) condition II to progress to the more severe condition III small break Loss of Coolant Accident (LOCA). To address the pressurizer water-solid concern for NPP Krsko RELAP5/MOD3.3 analyses of the LOOP event for best-estimate and USAR based scenarios have been performed. Different CVCS charging and letdown operation modes including the most conservative case with CVCS charging flow at maximum and letdown flow isolated were analyzed. (author)
[en] Highlights: • The Primary Heat Transfer System of the WCCB blanket for CFETR was designed and modeled using RELAP5 code. • Three typical accident cases, loss of flow accident, in-vessel loss of coolant accident and ex-vessel loss of coolant accident were simulated and analyzed. • No serious damage was caused during the loss of flow accident due to the establishment of natural circulation of coolant. • The integrity of the confinement barriers can be ensured for a limited period of time during loss of coolant accidents. • Additional safety facility are needed to prevent serious consequences and mitigation methods are discussed. - Abstract: The Water Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR). In this work, the Primary Heat Transfer System (PHTS) of the WCCB blanket was designed based on the configuration of the blanket sectors, employing two identical loops at this stage. Each loop consists of a steam generator, a pressurizer and two pumps, feeding water coolant into each blanket modules individually of 8 blanket sectors. One of the loop was modeled using RELAP5/MOD3.3 under normal condition and accident cases. The operational mode of PHTS was carefully chosen so as to obtain a more stable hydraulic behavior under steady state, due to the anisotropy of geometry structures and heat sources. Enveloping accidental cases, including Loss of Flow Accident (LOFA), in-vessel Loss of Coolant Accident (LOCA), and ex-vessel LOCA, were selected to preliminarily evaluate the safety performance of the system. The results show that the integrity of the confinement barriers can only be ensured in limited period of time during LOCAs. Additional safety facilities are needed and mitigation methods: are discussed.