Decay Heat
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Md Rezouanul Kabir ◽  
Morozov A.V. ◽  
Md Saif Kabir

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.

2021 ◽  
Vol 1 (2) ◽  
pp. 11-19
Catur Febriyanto Sutopo ◽  
Arifin M. Susanto

IN 2021, BAPETEN, AS THE REGULATORY BODY, IS ESTABLISHING A BAPETEN REGULATION REGARDING THE REACTOR COOLANT SYSTEM AND RELATED SYSTEMS, WHICH CURRENTLY ARE NOT YET AVAILABLE. Therefore, it is crucial to establish the BAPETEN Regulation. Based on the reasons, before setting the BAPETEN Regulation, it is necessary to conduct a study that is expected to provide a more comprehensive description and provide recommendations on what things need to be regulated in the BAPETEN Regulation, especially for gas-cooled reactors. The method used in this study is a literature study from various relevant references. The result of this study is that it is essential to require a capacity of the ultimate heat sink, including the spent nuclear fuel storage pool and a minimum period of the ability of the top heat sink in the accident analysis if the decay heat in the storage pool and the residual heat in the reactor core fail simultaneously. On the other hand, it is also necessary to require a margin of uncertainty to evaluate a situation and take corrective action. Likewise, independent and redundant access to the ultimate heat sink is needed to increase reliability. As for gas-cooled reactors, it is required to adapt the terms used. In addition, it is necessary to determine the appropriate definition because some of the terms used in water-cooled reactors have the same terms as gas-cooled reactors but have different functions. Keywords: Regulatory assessment, coolant system, related systems, gas-cooled reactors

Ganesh Vythilingam ◽  
Parimal Pramod Kulkarni ◽  
Arun Nayak

Abstract Some of the advanced nuclear reactors employ an ex-vessel core catcher to mitigate core melt scenarios by stabilizing and cooling the corium for prolonged period by strategically flooding it. The side indirect cooling with top flooding strategy described in this study may lead to water ingression either through the melt crust which may lead to interaction between un-oxidised metal in the melt and water leading to hydrogen production. In order to avoid this deleterious scenario, water ingression into the bulk of the melt should be avoided. The studies described in this manuscript show that water ingression depends on the flooding strategy, i.e. the time delay between top flooding and melt relocation. Two experiments under identical conditions of simulant temperature, melt material and test section geometry were conducted with simulated decay heat of 1 MW/m3. Sodium borosilicate glass was used as the corium simulant. In the first experiment, water was flooded onto the top of melt pool soon after melt relocation. In the second experiment, water flooding at the top of melt pool was made after 30 minutes of the melt relocation. The results show that a finite time delay of introduction of water onto the top of the melt pool is paramount to engender the development of a stable crust around the melt and therefore eliminating water ingression into melt pool and ensuring controlled coolability of the melt.

2021 ◽  
Vol 173 ◽  
pp. 112906
Lavinia Vicini ◽  
Nicolás Schiliuk ◽  
Eugenio Coscarelli ◽  
Giovanni Dell'Orco ◽  
Gianfranco Caruso

Energies ◽  
2021 ◽  
Vol 14 (23) ◽  
pp. 7862
Changhwan Lim ◽  
Jonghwi Choi ◽  
Hyungdae Kim

A fork-type heat pipe (FHP) is a passive heat-transport and air-cooling device used to remove the decay heat of spent nuclear fuels stored in a liquid pool during a station blackout. FHPs have a unique geometrical design to resolve the significant mismatch between the convective heat transfer coefficients of the evaporator and condenser parts. The evaporator at the bottom is a single heat-exchanger tube, whereas the condenser at the top consists of multiple finned tubes to maximize the heat transfer area. In this study, the heat transfer characteristics and operating limits of an FHP device were investigated experimentally. A laboratory-scale model of an FHP was manufactured, and a series of tests were conducted while the temperature was varied to simulate a spent fuel pool. As an index of the average heat transfer performance, the loop conductance was computed from the measurement data. The results show that the loop conductance of the FHP increased with the heat transfer rate but deteriorated significantly at the operating limit. The maximum attainable heat transfer rate of the unit FHP model was accurately predicted by the existing correlations of the counter-current flow limit for a single-rod-type heat pipe. In addition, the instant heat transfer behaviors of the FHP model under different temperature conditions were examined to interpret the measured loop conductance variation and operating limit.

2021 ◽  
pp. 104042
Justin B. Clarity ◽  
Henrik Liljenfeldt ◽  
Kaushik Banerjee ◽  
L. Paul Miller

Energies ◽  
2021 ◽  
Vol 14 (21) ◽  
pp. 6916
Francesco Galleni ◽  
Marigrazia Moscardini ◽  
Andrea Pucciarelli ◽  
Maria Teresa Porfiri ◽  
Nicola Forgione

This work presents a thermohydraulic analysis of a postulated accident involving the rupture of the breeder primary cooling loop inside a heat exchanger (once through steam generator). After the detection of the loss of pressure inside the primary loop, a plasma shutdown is actuated with a consequent plasma disruption, isolation of the secondary loop, and shutoff of the pumps in the primary; no other safety counteractions are postulated. The objective of the work is to analyze the pressurization of the primary and secondary sides to show that the accidental overpressure in the two sides of the steam generators is safely accommodated. Furthermore, the effect of the plasma disruption on the FW, in terms of temperatures, should be analyzed. Lastly, the time transients of the pressures and temperatures in the HX and BB for a time span of up to 36 h should be obtained to assess the effect of the decay heat over a long period. A full nodalization of the OTSG was realized together with a simplified nodalization of the whole PHTS BB loop. The code utilized was MELCOR for fusion version 1.8.6. The accident was simulated by activating a flow path which directly connected one section of the primary with the parallel section of the secondary side. It is shown here that the pressures and the temperatures inside the whole PHTS system remain below the safety thresholds for the whole transient.

Kerntechnik ◽  
2021 ◽  
Vol 86 (5) ◽  
pp. 338-342
R. David

Abstract During the in-vessel stage of a severe accident in a CANDU 6 reactor, decay heat from a collapsed core would be rejected through the calandria walls into the surrounding water. At the step in the calandria wall, the subshell and annular plate meet at a right angle pointing into the calandria. The geometry at this joint could concentrate the exiting heat flux, potentially leading to calandria failure. Finite element analysis is used to study the heat transfer near the welded joint. Different weld profiles, boundary conditions, and decay heat characteristics are considered, and the local concentration of exiting heat flux is calculated.

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