Options for management of Containment Integrity during severe accident in Indian PHWR

2017 ◽  
Vol 323 ◽  
pp. 386-393 ◽  
Author(s):  
Sanjeev Kr. Sharma ◽  
D.K. Bhartia ◽  
N. Mohan
Author(s):  
Taehoon Kim ◽  
Sukyoung Pak ◽  
Yongjin Cho

During a severe accident, contact of the molten corium with the coolant water may cause an energetic steam explosion which is a rapid increase of explosive vaporization by transfer to the water of a significant part of the energy in the corium melt. This steam explosion has been considered as an adverse effect when the water is used to cool the molten corium and could threaten reactor vessel, reactor cavity, containment integrity. In this study, TROI TS-2 and TS-3 experiments as part of the OECD/SERENA-2 project were analyzed with TEXAS-V. Input parameters were based on actual TROI experiment data. In mixing simulations, calculated results were compared to melt front behavior, void fraction in trigger time and other parameters in experiment results. In explosion simulations, corresponding to TROI experiments an external triggering was employed at the moment that melt front reached heights of 0.4 m. Calculated results of peak pressure and impulse at the bottom were compared with TROI experiment results. Melt front behaviors of the melt was different from the experimental results in both TS-2 and TS-3. Void fraction in triggering time in TS-2 was in good agreement with the experiment results and in TS-3 was slightly overestimated. The peak pressure and impulse at bottom were successfully predicted by TEXAS-V. These calculations will allow establishing whether the limitations and differences observed in the simulations of the experiments are important for the reactor case.


2011 ◽  
Vol 2011 ◽  
pp. 1-13 ◽  
Author(s):  
Stéphane Mimouni ◽  
Namane Mechitoua ◽  
Mehdi Ouraou

A large amount of Hydrogen gas is expected to be released within the dry containment of a pressurized water reactor (PWR), shortly after the hypothetical beginning of a severe accident leading to the melting of the core. According to local gas concentrations, the gaseous mixture of hydrogen, air and steam can reach the flammability limit, threatening the containment integrity. In order to prevent mechanical loads resulting from a possible conflagration of the gas mixture, French and German reactor containments are equipped with passive autocatalytic recombiners (PARs) which preventively oxidize hydrogen for concentrations lower than that of the flammability limit. The objective of the paper is to present numerical assessments of the recombiner models implemented in CFD solvers NEPTUNE_CFD and Code_Saturne. Under the EDF/EPRI agreement, CEA has been committed to perform 42 tests of PARs. The experimental program named KALI-H2, consists checking the performance and behaviour of PAR. Unrealistic values for the gas temperature are calculated if the conjugate heat transfer and the wall steam condensation are not taken into account. The combined effects of these models give a good agreement between computational results and experimental data.


Author(s):  
Jianjun Wang ◽  
Beibei Luo ◽  
Xueqing Guo ◽  
Zhongning Sun

In this paper, a closed loop concept, which is composed of two heat exchangers with same scale, pipes, valves and one tank, has been developed as a passive containment cooling system for a large dry concrete containment. The system is designed to maintain the containment integrity by taking the heat drained into the containment following a severe accident, e.g. LOCA or MSLB. Under different conditions in containment, the system may operate in single phase mode or two phase mode. According to the design limitation of containment, the fixed temperature boundary condition is applied to the system analysis. We have developed the codes for the analysis of the system by ourselves. The operating behaviors of the system are studied numerically from startup to long term operation. In the light of the fact that the fraction of steam in the containment may be changing during the accident scenario, it is reasonable that the heat transfer coefficient will be different. Therefore, the sensitivity analysis of the heat transfer coefficient is also performed. Based on the results and corresponding analysis, it can be concluded that the system may be utilized to meet the design purpose for the containment integrity requirement.


Author(s):  
Zhiqiang Zou ◽  
Jian Deng ◽  
Yu Zhang

Hydrogen combustion in the containment building may be a threat to containment integrity. Especially in some subcompartments, local hydrogen detonation likely to happen, which may destroy the containment structure or the safety equipments in the containment. A study has been carried out using the GASFLOW three-dimensional CFD code to evaluate the hydrogen distribution in the subcompartment during a severe accident. The GASFLOW calculation has provided detailed results for the spatial distribution of gas concentrations in the reactor coolant pump (RCP) compartment and the steam generator (SG) compartment within containment, the high local hydrogen concentration was found in the local area. Then, three kinds of different hydrogen mitigation measures which were optimizing compartment structure, hydrogen recombiner and igniter were analyzed and compared. The σ criterion and λ criterion are used for conservative estimates of the flame acceleration (FA) potential and the possibility of deflagration to detonation transition (DDT) in the compartment after using the mitigation measures, respectively.


Author(s):  
A. C. Morreale ◽  
L. S. Lebel ◽  
M. J. Brown

Severe accidents are of increasing concern in the nuclear industry worldwide since the accidents at Fukushima Daiichi (March 2011). These events have significant consequences that must be mitigated to ensure public and employee safety. Filtered containment venting (FCV) systems are beneficial in this context as they would help to maintain containment integrity while also reducing radionuclide releases to the environment. This paper explores the degree to which filtered containment venting would reduce fission product releases during two Canada Deuterium Uranium (CANDU) 6 severe accident scenarios, namely a station blackout (SBO) and a large loss of coolant accident (LLOCA) (with limited emergency cooling). The effects on the progression of the severe accident and radionuclide releases to the environment are explored using the Modular Accident Analysis Program (MAAP)–CANDU integrated severe accident analysis code. The stylized filtered containment venting system model employed in this study avoids containment failure and significantly reduces radionuclide releases by 95–97% for non-noble gas fission products. Filtered containment venting is shown to be a suitable technology for the mitigation of severe accidents in CANDU, maintaining containment integrity and reducing radionuclide releases to the environment.


Author(s):  
Khurram Mehboob ◽  
Xinrong Cao ◽  
Majid Ali ◽  
Rehan Khan

Since, containment integrity is the main issue under accidental conditions. Radiological consequences of LWR under accident have the grievous impact on the reactor building and its surrounding environment. Iodine is one of the most hazardous fission product releases in the serious accidents. So in this paper, the iodine source term has been evaluated for two-loop PWR under severe accident initiated due to LOCA. The TMI-2 reactor is considered as the reference reactor. The modeling and simulations are carried out by developing a MATLAB base program that uses the post-accident conditions and core inventory as input. The containment response, in order to mitigate the environmental and in-containment iodine source term is studied in normal, emergency, and isolation states of containment. The In-containment iodine source term is calculated with, and without the operation of engineering safety features (ESFs). The mitigation is determined by the activation of ESF. The environmental iodine source term is calculated as the function of containment response. The iodine dependency on the containment retention factor is also studied in all said states of containment. Results indicate the weak sensitivity of Iodine with activation of ESF towards exhaust rate values, under ESFs Operation.


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.


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):  
Deng Jian ◽  
Xuewu Cao

Various studies have shown that hydrogen combustion is one of major risk contributors to threaten the integrity of the containment in a nuclear power plant. That hydrogen risk should be considered in severe accident strategies in current and future NPPs has been emphasized in the latest policies issued by the National Nuclear Safety Administration of China (NNSA). According to a deterministic approach, three typical severe accident sequences for a PWR large dry containment, such as the large break loss-of-coolant (LLOCA), the station blackout (SBO), and the small break loss-of-coolant (SLOCA) are analyzed in this paper with MELCOR code. Hydrogen concentrations in different compartments are observed to evaluate the potential hydrogen risk. The results show that there is a great amount of hydrogen released into the containment, which causes the containment pressure to increase and some potential inconsecutive burnings. Therefore, certain hydrogen management strategies should be considered to reduce the risk to threaten the containment integrity.


2014 ◽  
Vol 2014 ◽  
pp. 1-10 ◽  
Author(s):  
Sang-Won Lee ◽  
Tae-Hyub Hong ◽  
Yu-Jung Choi ◽  
Mi-Ro Seo ◽  
Hyeong-Taek Kim

After the Fukushima Daiichi nuclear power plant accident, the Korean government and nuclear industries performed comprehensive safety inspections on all domestic nuclear power plants against beyond design bases events. As a result, a total of 50 recommendations were defined as safety improvement action items. One of them is installation of a containment filtered venting system (CFVS) or portable backup containment spray system. In this paper, the applicability of CFVS is examined for OPR1000, a 1000 MWe PWR with large dry containment in Korea. Thermohydraulic analysis results show that a filtered discharge flow rate of 15 [kg/s] at 0.9 [MPa] is sufficient to depressurize the containment against representative containment overpressurization scenarios. Radiological release to the environment is reduced to10-3considering the decontamination factor. Also, this cyclic venting strategy reduces noble gas release by 50% for 7 days. The probability of maintaining the containment integrity in level 2 probabilistic safety assessment (PSA) initiating events is improved twofold, from 43% to 87%. So, the CFVS can further improve the containment integrity in severe accident conditions.


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