Retention of Molten Corium in Calandria Vessel of PHWR With Moderator Drain Pipe

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Pradeep Pandey ◽  
Parimal P. Kulkarni ◽  
Arun Nayak ◽  
Sumit V. Prasad

Abstract Retention of molten corium inside calandria vessel is crucial for arresting accident progression in pressurized heavy water reactors (PHWRs) during severe accidents. Our earlier tests have demonstrated corium retention and its cooling inside the calandria vessel of PHWRs through external cooling by vault water. However, the presence of nozzles and moderator drain pipe at the bottom of calandria vessel has not been considered in these studies. These nozzles and drain pipes used for moderator circulation can make the viability of corium retention even more challenging. Once the moderator has evaporated, debris reheating, compacting, and finally melting can cause the release of molten corium into the moderator recirculation system. This can lead to the relocation of corium beyond calandria vessel. The corium might reach the pump room or calandria vault after the failure of moderator drain pipe and/or moderator pump seals. This has severe consequences on containment integrity due to molten corium concrete interaction (MCCI). The risks posed by MCCI can be avoided if corium can be contained inside calandria vessel even with the presence of nozzles (at the bottom of the vessel) or if at all it enters into the drain line, does not cause its failure. Thus, it becomes crucial to evaluate the challenges faced by “in-vessel retention” (IVR) as a severe accident management strategy due to the presence of openings in the calandria vessel. Relatively colder debris present near the bottom of calandria vessel might help in obstructing the nozzles of the moderator drain line and can prevent the entry of hot molten corium into the moderator cooling line. The role of debris, therefore, becomes important under such scenarios for not just insulation of calandria vessel from hot corium but also for retention of corium within the vessel. In this article, these issues are addressed by conducting two sets of experiments for assessment of retention capability (IVR) of calandria vessel: (i) with the presence of debris and (ii) without debris at the bottom of calandria vessel. The moderator recirculation line was scaled to simulate the heat transfer from corium to vault water and solidification of corium simulant while flowing through the moderator drain pipe. It was observed that debris bed present at the bottom of the vessel helps in arresting the molten corium front and thus prevents corium from entering into moderator drain pipe. When experiments were conducted without debris, molten corium was found to be relocating in the moderator drain pipe. The drain pipe, however, did not fail under the thermal load.

Author(s):  
Nikolay Ivanov Kolev

This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have been determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool.


1994 ◽  
Vol 106 (2) ◽  
pp. 139-154 ◽  
Author(s):  
Donghan Yu ◽  
Leiming Xing ◽  
William E. Kastenberg ◽  
David Okrent

Author(s):  
Junrong Wang ◽  
Huajian Chang ◽  
Wenxiang Zheng ◽  
Zhiwei Zhou

In-vessel retention (IVR) of core melt through external reactor vessel cooling (ERVC) is a key severe accident management strategy to ensure that the vessel head remains intact and eliminate consequent major threats to containment integrity. To maintain the margin against the failure of the reactor vessel and make sure of the feasibility of IVR for its role to confine the molten corium with 1400MW power, CAP1400, Chinese version of large passive PWR, systematic investigations including experimental and analytical researches of IVR are very important to the development of CAP1400. This paper briefly reviews the progress and tasks of a four-year project which was planned to analyze, evaluate, improve and validate the effectiveness of IVR employed in the CAP1400.


2019 ◽  
Vol 2019 ◽  
pp. 1-18
Author(s):  
Sergey Galushin ◽  
Pavel Kudinov

Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.


2019 ◽  
Vol 7 (2B) ◽  
Author(s):  
Seung Min Lee ◽  
Nelbia Da Silva Lapa ◽  
Gaianê Sabundjian

The aim of this work was to simulate a severe accident at a typical PWR, initiated with a break in Emergency Core Cooling System line of a hot leg, using the MELCOR code. The model of this typical PWR was elaborated by the Global Research for Safety and provided to the CNEN for independent analysis of the severe accidents at Angra 2, which is similar to this typical PWR. Although both of them are not identical, the results obtained of that typical PWR may be valuable because of the lack of officially published simulation of severe accident at Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes, after the break at the hot leg, were calculated as well as degree of core degradation and hydrogen production within the containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management by implementing each measure in this model.


2019 ◽  
Vol 2019 ◽  
pp. 1-10 ◽  
Author(s):  
Hao Yu ◽  
Minjun Peng

Interest in evaluation of severe accidents induced by extended station blackout (ESBO) has significantly increased after Fukushima. In this paper, the severe accident process under the high and low pressure induced by an ESBO for a small integrated pressurized water reactor (IPWR)-IP200 is simulated with the SCDAP/RELAP5 code. For both types of selected scenarios, the IP200 thermal hydraulic behavior and core meltdown are analyzed without operator actions. Core degradation studies firstly focus on the changes in the core water level and temperature. Then, the inhibition of natural circulation in the reactor pressure vessel (RPV) on core temperature rise is studied. In addition, the phenomena of core oxidation and hydrogen generation and the reaction mechanism of zirconium with the water and steam during core degradation are analyzed. The temperature distribution and time point of the core melting process are obtained. And the IP200 severe accident management guideline (SAMG) entry condition is determined. Finally, it is compared with other core degradation studies of large distributed reactors to discuss the influence of the inherent design characteristics of IP200. Furthermore, through the comparison of four sets of scenarios, the effects of the passive safety system (PSS) on the mitigation of severe accidents are evaluated. Detailed results show that, for the quantitative conclusions, the low coolant storage of IP200 makes the core degradation very fast. The duration from core oxidation to corium relocation in the lower-pressure scenario is 53% faster than that of in the high-pressure scenario. The maximum temperature of liquid corium in the lower-pressure scenario is 134 K higher than that of the high-pressure scenario. Besides, the core forms a molten pool 2.8 h earlier in the lower-pressure scenario. The hydrogen generated in the high-pressure scenario is higher when compared to the low-pressure scenario due to the slower degradation of the core. After the reactor reaches the SAMG entry conditions, the PSS input can effectively alleviate the accident and prevent the core from being damaged and melted. There is more time to alleviate the accident. This study is aimed at providing a reference to improve the existing IPWR SAMGs.


Author(s):  
Longze Li ◽  
Jue Wang ◽  
Yapei Zhang ◽  
G. H. Su

The natural circulation small modular reactor (NCSMR) is a 330 MW reactor which has no reactor coolant pumps (RCP) and no active safety injection systems at all. The reactor is mainly comprised of the reactor pressure vessel (RPV) with integral pressurize r and steam generator. RPV is enclosed by a vacuumed pressure containment vessel (PCV) and the PCV is submerged in the underground containment pool. A MELCOR model and corresponding input deck are developed for the RPV, PCV, and containment pool. The containment pool takes the role of ultimate heat sink (UHS) in accident situations. The containment pool may crack and leak in some critical accidents as the earthquake, leading to the severe accident of the reactor. A TMI-2 like SBLOCA in the RPV (stuck open RVVs) along with the containment pool crack (loss of ultimate heat sink) is simulated in the work. So me key parameters as the RRVs stuck open fraction, the PCV-SRVs open or not, the containment pool crack position would have large influence on the severe accident sequence. The sensitivity of these parameters to the accident sequence is analyzed in the work. According to the simulation results, the RPV pressure decreased with the RRVs stuck open. The depressurization of RPV accelerated with the RPV-SRV open fraction increase. The PCV pressure increased after that. Two cases as the PCV-SRV open after PCV pressure increase to 5 MPa, and PCV break while the RV d id not open, are analysis. The coolant discharge mass flo wrate in RPV and PCV were different in two cases, leading to the different degradation situation of the core. Since the containment pool is so important for the accident mitigation, sensitivity analysis is done for the containment pool crack position in the pool. The work will be meaningful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents during the design process of NCSMR.


Author(s):  
Tamás János Katona ◽  
András Vilimi

Paks Nuclear Power Plant identified the post-Fukushima actions for mitigation and management of severe accidents caused by external events that include updating of some hazard assessments, evaluation of capacity / margins of existing severe accident management facilities, and construction of some mew systems and facilities. In all cases, the basic question was, what level of margin has to be ensured above design basis external hazard effects, and what level of or hazard has to be taken for the design. Paks Nuclear Power Plant developed certain an applicable in the practice concept for the qualification of already implemented and design the new post-Fukushima measures that is outlined in the paper. The concept and practice is presented on several examples.


2005 ◽  
Vol 152 (3) ◽  
pp. 253-265 ◽  
Author(s):  
Te-Chuan Wang ◽  
Shih-Jen Wang ◽  
Jyh-Tong Teng

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