Effect of Boric Acid Solubility in Steam on the Process of Mass Transfer during Emergency Cooling of VVER-1200 Nuclear Reactor

The mechanisms of boric acid mass transfer in a VVER-1200 reactor core are studied in this work in the event of a major circulatory pipeline rupture and loss of all AC power. The VVER-1200's passive core cooling technology is made up of two levels of hydro accumulators. They use boric acid solution with a concentration of 16 g H3BO3/kg H2O to control the reactivity. Because of the long duration of the accident process, the coolant with high boron content starts boiling and steam with low concentration of boric acid departs the core. So, conditions could arise in the reactor for possible accumulation and subsequent crystallization of boric acid, causing the core heat removal process to deteriorate. Calculations were carried out to estimate the likelihood of H3BO3 build-up and subsequent crystallization in the core of the VVER reactor. According to the calculations, during emergency the boric acid concentration in the reactor core is 0.153 kg/ kg and 0.158 kg/kg in both the events of solubility of steam and without solubility of steam respectively and it does not exceed the solubility limit which is about 0.415 kg/kg at water saturation temperature. No precipitation of boric acid occurs within this time during the whole emergency process. Therefore, findings of the study can be used to verify whether the process of decay heat removal is affected or not.

Simulation of VVER-1000 Guillotine Large Break Loss of Coolant Accident Using RELAP5/SCDAPSIM/MOD3.5

The safety performance of nuclear power plants (NPPs) is a very important factor in evaluating nuclear energy sustainability. Safety analysis of passive and active safety systems have a positive influence on reactor transient mitigation. One of the common transients is primary coolant leg rupture. This study focused on guillotine large break loss of coolant (LB-LOCA) in one of the reactor vessels, in which cold leg rupture occurred, after establishment of a steady-state condition for the VVER-1000. The reactor responses and performance of emergence core cooling systems (ECCSs) were investigated. The main safety margin considered during this simulation was to check the maximum value of the clad surface temperature, and it was then compared with the design licensing limit of 1474 K. The calculations of event progression used the engineering-level RELAP5/SCDAPSIM/MOD3.5 thermal-hydraulic program, which also provide a more detailed treatment of coolant system thermal hydraulics and core behavior. The obtained results show that actuation of ECCSs at their actuation set points provided core cooling by injecting water into the reactor pressure vessel, as expected. The peak cladding temperature did not overpass the licensing limit during this LB-LOCA transient. The primary pressure above the core decreased rapidly from 15.7 MPa to 1 MPa in less than 10 s, then stabilizes up to the end of transient. The fuel temperature decreased from 847 K to 378 K during the first 30 s of the transient time. The coolant leakage reduced from 9945 kg/s to approximately 461 kg/s during the first 190 s in the transient. Overall, the study shows that, within the design of the VVER-1000, safety systems of the have inherent robustness of containing guillotine LB-LOCA.

Analysis of IP200 Severe Accident Process Response to SBO and Emergency Power Failure

Advanced SMRs such as the integrated pressurized water reactor IP200 use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In this thesis, the IP200 severe accident induced by the SBO and emergency power failure was modeled and analyzed using RELAP5 / SCDAP / MOD3.4 code. Based on the steady state calculation, which agrees well with designed values, the SBO accident for transient calculation is carried out. First, the case of the SBO accident without the passive core cooling system was calculated. The progression and scenario in the RPV was simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. Then, mitigation measures PRHRS and CMT were put in at four different time points when the core is began to uncovered, the core is completely uncovered, hydrogen is began to produced, and the molten pool is formed. The results show that putting in mitigation measures before the accident progresses to the point where the core starts to produce hydrogen can ensure that the core does not melt and avoid hydrogen risk.

Experimental evaluation of natural convection in the IPR-R1 Triga research reactor at 264 kW and 105 kW

In order to study the safety aspects connected with the permanent increase of the maximum steady state power of the IPR-R1 Triga Reactor of the Nuclear Technology Development Center (CDTN), experimental measurements were done with the reactor operating at power levels of 265 kW and 105 kW, with the pool forced cooling system turned off. A number of parameters were measured in real-time such as fuel and water temperatures, radiation levels, reactivity, and influence of cooling system. Information on all aspects of reactor operation was displayed on the Data Acquisition System (DAS) shown the IPR-R1 online performance. The DAS was developed to monitor and record all operational parameters. Information displayed on the monitor was recorded on hard disk in a historical database. This paper summarizes the behavior of some operational parameters, and in particular, the evolution of the temperature in the fuel element centerline positioned in the core hottest location. The natural circulation test was performed to confirm the cooling capability of the natural convection in the IPR-R1 reactor. It was confirmed that the IPR-R1 has capability of long-term core cooling by natural circulation operating at 265 kW. The measured maximum fuel temperature of about 300 oC was lower than the operating limit of 550 oC. It has been proven that without cooling in the primary the gamma dose rate above reactor pool at high power levels was rather high.

Evaluation of Passive Safety Injection System Performance Under Large Break LOCA for Qinshan PWR

In this work, a brand new passive safety injection system has been designed for the ocean-based Qinshan Phase I nuclear power plant to update and replace the traditional active ones. The passive safety injection system is made up of high pressure, medium pressure, lower pressure safety injection system, and a two-stage automatic depressurization system. To evaluate the safety injection system performance, double-ended cold leg large break LOCA has been analyzed by best-estimated safety analysis RELAP5 code. The main operation and safety parameters such as primary system pressure, safe injection mass flow rates, core water level, and peak cladding temperature have been presented. The results conclude that the safety injection system can act as similar to that of the AP1000, which can assure sufficient core cooling and keep the reactor covered by the cold water under the most severe LBLOCA condition.

Investigation of effects of nonavailability of passive safety systems on the reactor behaviour during LOCA Scenario in AP600

The AP600 is a Westinghouse Advanced Passive PWR with a two–loop 1 940 MWt. This reactor is equipped with advanced passive safety systems which are designed to operate automatically at desired set-points. On the other hand, the failure or nonavailability to operate of any of the passive safety systems may affect reactor safety. In this study, modeling and nodalization of primary and secondary loops, and all passive reactor cooling systems are conducted and a 10-inch cold leg break LOCA is analyzed using ATHLET 3.1A Code. During loss of coolant accident in which the passive safety system failure or nonavailability are considered, four different scenarios are assumed. Scenario 1 with the availability of all passive systems, scenario 2 is failure of one of the accumulators to activate, scenario 3 is without actuation of the automatic depressurization system (ADS) stages 1–3, and scenario 4 is without actuation of ADS stage 4. Results indicated that the actuation of passive safety systems provide sufficient core cooling and thus could mitigate the accidental consequence of LOCAs. Failure of one accumulator during LOCA causes early actuation of ADS and In-Containment Refueling Water Storage Tank (IRWST). In scenario 3 where the LOCA without ADS stages 1–3 actuations, the depressurization of the primary system is relatively slow and the level of the core coolant drops much earlier than IRWST actuation. In scenario 4 where the accident without ADS stage-4 activation, results in slow depressurization and the level of the core coolant drops earlier than IRWST injection. During the accident process, the core uncovery and fuel heat up did not happen and as a result the safety of AP600 during a 10-in. cold leg MBLOCA was established. The relation between the cladding surface temperature and the primary pressure with the actuation signals of the passive safety systems are compared with that of RELAP5/Mode 3.4 code and a tolerable agreement was obtained.

First palaeointensity data from the cryogenian and their potential implications for inner core nucleation age

SUMMARY The timing of inner core nucleation is a hugely significant event in Earth's evolution and has been the subject of intense debate. Some of the most recent theoretical estimates for the age of nucleation fall throughout the Neoproterozoic era; much younger than previously thought. A young inner core requires faster recent core cooling rates and a likely hotter early core; knowledge of its age would be invaluable in understanding Earth's thermal history and total energy budget. Predictions generated by numerical dynamo models need to be tested against such data, but records are currently much too sparse to constrain the event to a precise period of time. Here, we present results from 720 Ma dolerite dykes (and one sill) from the Franklin Large Igneous Province, which fall within a crucial 300 Myr gap in palaeointensity records. This study uses three independent techniques on whole rocks from 11 sites spread across High Arctic Canada and Greenland to produce virtual dipole moments ranging from 5 to 20 ZAm2 (mean 11 ZAm2); almost one order of magnitude lower than the present-day field. These weak-field results agree with recent ultralow palaeointensity data obtained from Ediacaran rocks formed ∼150 Myr later and may support that the dynamo was on the brink of collapse in the Neoproterozoic prior to a young inner core formation date.