scholarly journals Radionuclide Inventory and Decay Heat Quantification Methodology for Severe Accident Simulations.

2014 ◽  
Author(s):  
Jeffrey Cardoni
Keyword(s):  
Severe Accident ◽  
Decay Heat ◽  
Radionuclide Inventory

Heat transfer at the step of a CANDU calandria during a severe accident

Kerntechnik ◽  
2021 ◽  
Vol 86 (5) ◽  
pp. 338-342
Author(s):  
R. David
Keyword(s):  
Heat Transfer ◽  
Finite Element Analysis ◽  
Heat Flux ◽  
Annular Plate ◽  
Welded Joint ◽  
Severe Accident ◽  
Local Concentration ◽  
Element Analysis ◽  
Surrounding Water ◽  
Decay Heat

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.

Experimental Study on Melt Coolability Capability of Calandria Vault Water During Severe Accident in Indian PHWRs for Prolonged Duration

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Heat Transfer ◽  
Single Phase ◽  
Convection Heat Transfer ◽  
Severe Accident ◽  
Transfer Coefficients ◽  
Decay Heat ◽  
Prolonged Duration ◽  
Water Reactors ◽  
Vessel Integrity ◽  
Inner Vessel

The present experimental investigation in a scaled facility of an Indian pressurized heavy water reactors (PHWRs) is focused on the heat transfer behavior from the calandria vessel (CV) to the calandria vault during a prolonged severe accident condition in the presence of decay heat. The transient heat transfer simulates the conditions from single phase to boiling in the calandria vault water, partial uncovery of the CV due to boil off of water in the vault, and refill of calandria vault. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics similar to prototypic material. About 60 kg of the molten material was poured into the test section at about 1100 °C. Decay heat in the melt pool was simulated by using high watt cartridge type heaters. The temperature distributions inside the molten pool across the CV wall thickness and vault water were measured for prolonged period which can be divided into various phases, viz., single phase natural convection heat transfer in calandria vault, boiling heat transfer in calandria vault, partial uncovery of CV, and refilling calandria vault. Experimental results showed that once the crust formed, the inner vessel temperature remained very low and vessel integrity maintained. Even boiling of calandria vault water and uncovery of CV had negligible effect on melt, CV, and vault water temperature. The heat transfer coefficients on outer vessel surface were obtained and compared with various conditions.

Evaluation of Dump Tank Coolability in PHWRs During Late-Phase Severe Accident

2019 ◽  
Vol 5 (4) ◽  
Author(s):  
Pradeep Pandey ◽  
Parimal P. Kulkarni ◽  
Arun Nayak ◽  
Sumit V. Prasad
Keyword(s):  
Heavy Water ◽  
Late Phase ◽  
Atomic Power Station ◽  
Severe Accident ◽  
Large Temperature ◽  
Tank Wall ◽  
Power Station ◽  
Melt Temperature ◽  
Decay Heat ◽  
Water Reactors

In some of the older design of pressurized heavy water reactors (PHWRs), such as in Rajasthan Atomic Power Station (RAPS) and Madras Atomic Power Station (MAPS), in case of a severe accident, the debris/corium may cause failure of the dump port of calandria and relocate into the dump tank. The sensible and decay heat of debris/corium makes the heavy water in dump tank to boil off leaving the dry debris in dump tank. The dry debris remelt with time and the molten corium, thus, formed has the potential to breach the dump tank and move into the containment cavity, which is highly undesirable. Hence, as an accident management strategy, water is being flooded outside the dump tank using fire water hook-up lines to remove the heat from corium to cool and stabilize it and terminate the accident progression, similar to in vessel retention. However, the question is “is the molten corium coolable by this technique.” The coolability of the molten corium in dump tank as in the reactor is assessed by conducting experiments in a scaled facility using a simulant material having comparable thermophysical properties with that of corium. Melting of dry debris resting on dump tank bottom marks the beginning of the experimental investigation for present analysis. Decay heat is simulated by a set of immersed heaters inside the melt. Temperature profiles at different locations in dump tank and in the melt pool are obtained as a function of time to demonstrate the coolability with decay heat. Large temperature gradient is observed within the corium, involving high melt center temperature and low tank wall temperature suggesting formation of crust which insulates the dump tank wall from hot corium.

Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code

2014 ◽  
Author(s):  
Zhang Dabin ◽  
Zhiwei Zhou ◽  
Heng Xie ◽  
Tang Yang
Keyword(s):  
Cooling System ◽  
Safety Performance ◽  
Severe Accident ◽  
Hybrid Reactor ◽  
Accident Analysis ◽  
Analysis Model ◽  
Decay Heat ◽  
Calculation Results ◽  
Core Meltdown ◽  
Dry Out

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily. Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.

Performance Evaluation of DRACS System for FHTR and Time Assessment of Operation Procedure

2013 ◽  
Author(s):  
Junya Nakata ◽  
Mikihiro Wakui ◽  
Michitsugu Mori ◽  
Hiroto Sakashita ◽  
Charles Forsberg
Keyword(s):  
Performance Evaluation ◽  
High Temperature ◽  
Cooling System ◽  
Heat Removal ◽  
Severe Accident ◽  
Air Cooling ◽  
Fluoride Salt ◽  
Nuclear Power Reactor ◽  
Decay Heat ◽  
Operation Procedure

The Fluoride-salt-cooled High-temperature Reactor (FHR) is a new concept of nuclear power reactor being investigated mainly in U.S. and China. The coolant is a liquid salt with a melting point of about 460°C and a boiling point of over 1400°C. As the baseline decay heat removal system, a passive Direct Reactor Air Cooling System (DRACS) is utilized. Though DRACS system has been developed in Sodium Fast reactors (SFR), there are some differences between both. For example, the system in FHR must decrease heat removal when temperatures are low to avoid freezing of the salt and blocking the flow of liquid. Therefore, considering its characteristics, a numerical investigation of DRACS system is needed to evaluate whether FHR has proper ability to remove decay heat and to be robust for a long-time cooling operation after even a severe accident. Furthermore, in addition to its performance evaluation, it is required to make up the operation plan of FHR considering features of this system. It is highly important, with the view of avoiding severe accident, to determine by when the system should be started up. In both countries mentioned above, Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently in progress to build. Reviewing its design and system is a crucial step needed to develop the FHR technology. In this research, a performance of DRACS system under some thermal-hydraulic basic events was evaluated by numerical simulation. This paper also suggested the adequate operation procedure suitable for FHTR to avoid a severe accident.

Study on Cooling Process in a Reactor Vessel of Sodium-Cooled Fast Reactor Under Severe Accident: Velocity Measurement Experiments Simulating Operation of Decay Heat Removal Systems

2020 ◽  
Author(s):  
Mitsuyo Tsuji ◽  
Kosuke Aizawa ◽  
Jun Kobayashi ◽  
Akikazu Kurihara ◽  
Yasuhiro Miyake
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Piv Measurement ◽  
Image Velocimetry ◽  
Experimental Apparatus

Abstract In Sodium-cooled Fast Reactors (SFRs), it is important to optimize the design and operate decay heat removal systems for safety enhancement against severe accidents which could lead to core melting. It is necessary to remove the decay heat from the molten fuel which relocated in the reactor vessel after the severe accident. Thus, the water experiments using a 1/10 scale experimental apparatus (PHEASANT) simulating the reactor vessel of SFR were conducted to investigate the natural circulation phenomena in a reactor vessel. In this paper, the natural circulation flow field in the reactor vessel was measured by the Particle Image Velocimetry (PIV) method. The PIV measurement was carried out under the operation of the dipped-type direct heat exchanger (DHX) installed in the upper plenum when 20% of the core fuel fell to the lower plenum and accumulated on the core catcher. From the results of PIV measurement, it was quantitatively confirmed that the upward flow occurred at the center region of the lower and the upper plenums. In addition, the downward flows were confirmed near the reactor vessel wall in the upper plenum and through outermost layer of the simulated core in the lower plenum. Moreover, the relationship between the temperature field and the velocity field was investigated in order to understand the natural circulation phenomenon in the reactor vessel. From the above results, it was confirmed that the natural circulation cooling path was established under the dipped-type DHX operation.

iB1350: Part 2 — Level1 PRA Considering Optimization of Safety Systems for the iB1350

2018 ◽  
Author(s):  
Tanaka Go ◽  
Sato Takashi ◽  
Komori Yuji ◽  
Matsumoto Keiji
Keyword(s):  
Cooling System ◽  
Heat Removal ◽  
Lessons Learned ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Active Safety ◽  
Decay Heat ◽  
Active Safety Systems ◽  
Fukushima Daiichi ◽  
Safety Systems

iB1350 stands for an innovative, intelligent and inexpensive BWR 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident, and has incorporated both the lessons learned from the Fukushima Daiichi accident and the WENRA safety objectives. It has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The IDHS currently consists of 4 division active safety systems for a DBA, and 2 division active safety systems as well as built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and an innovative passive containment cooling system (iPCCS) for a Severe Accident (SA), which brings the total to 6 division active safety systems. Taking into account of excellent feature of the BiPSS, the IDHS has potential to optimize its 6 division active safety systems. The iPCCS that composes the BiPSS has been enhanced and has greater capability to remove decay heat than the conventional PCCS. While the conventional PCCS never can cool the S/P, the iPCCS can automatically cool the S/P directly with benefits from the structure of the Mark W containment. That makes it possible for the iB1350 to cool the core using only core inject systems and the iPCCS without RHR system: conventional active decay heat removal system. To make the most of this excellent feature of the iPCCS, it is under consideration to take credit for the iPCCS as safety systems for a DBA to optimize configuration of the IDHS. Currently, there are several proposed configurations of the IDHS that are expected to achieve cost reduction and enhance its reliability resulting from passive feature of the iPCCS. To compare those configurations of the IDHS, Level 1 Internal Events Probabilistic Risk Assessment (PRA) and sensitivity analyses considering external hazards have been performed for each configuration to provide measure of plant safety.

In-Calandria Retention of Corium in PHWR: Experimental Investigation and Remaining Issues

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Sumit V. Prasad ◽  
A. K. Nayak
Keyword(s):  
Experimental Investigation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Heat Removal ◽  
Public Domain ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Decay Heat ◽  
Critical Issues

After the Fukushima accident, the public has expressed concern regarding the safety of nuclear power plants. This accident has strengthened the necessity for further improvement of safety in the design of existing and future nuclear power plants. Pressurized heavy water reactors (PHWRs) have a high level of defense-in-depth (DiD) philosophy to achieve the safety goal. It is necessary for designers to demonstrate the capability of decay heat removal and integrity of containment in a PHWR reactor for prolonged station blackout to avoid any release of radioactivity in public domain. As the design of PHWRs is distinct, its calandria vessel (CV) and vault cooling water offer passive heat sinks for such accident scenarios and submerged calandria vessel offers inherent in-calandria retention (ICR) features. Study shows that, in case of severe accident in PHWR, ICR is the only option to contain the corium inside the calandria vessel by cooling it from outside using the calandria vault water to avoid the release of radioactivity to public domain. There are critical issues on ICR of corium that have to be resolved for successful demonstration of ICR strategy and regulatory acceptance. This paper tries to investigate some of the critical issues of ICR of corium. The present study focuses on experimental investigation of the coolability of molten corium with and without simulated decay heat and thermal behavior of calandria vessel performed in scaled facilities of an Indian PHWR.

Numerical Simulation of the Behavior of Fission Products in the Primary Circuit of the VVER During the LOCA Severe Accident

2009 ◽  
Author(s):  
S. V. Tsaun ◽  
V. V. Bezlepkin ◽  
A. E. Kiselev ◽  
I. A. Potapov ◽  
V. F. Strizhov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Fission Products ◽  
Severe Accident ◽  
Primary Circuit ◽  
Decay Heat ◽  
The Core ◽  
Severe Accidents ◽  
Loss Of Coolant ◽  
Mass Activity

The methods and models for the analysis of the radiological consequences of the design basis and severe accidents in a Nuclear Power Plant (NPP) are presented in this paper when using the system code SOCRAT. The system code SOCRAT/V3 was elaborated for a realistic analysis of radiological consequences of severe accidents in a NPP. The following models of the fission products (FP) behavior are included into the code SOCRAT/V3: (i) the condensation and the evaporation of the FP in the gaseous phase and (ii) the sedimentation, the evaporation, and the coagulation of the aerosol-shape FP. The latter processes are governed by gravity, Brownian and turbulent diffusion, thermophoresis, turbophoresis and so forth. The behavior of the FP during the loss-of-coolant accidents (LOCA) is presented to demonstrate the possibilities of the code SOCRAT/V3. The main stages of the accident (the core dryout, the core reflooding, the core degradation, the hydrogen generation, the FP release, etc.) are described. Corresponding estimations of the mass, activity, and decay heat of the suspended, settled and released into containment the FP (Xe, Te, Cs, CsI, Cs2MoO4, and so forth) are represent as well.

Review on Water Radiolysis in the Fukushima Daiichi Accident: Potential Cause of Hydrogen Generation and Explosion

2014 ◽  
Author(s):  
Genn Saji
Keyword(s):  
Hydrogen Generation ◽  
Hydrogen Gas ◽  
Severe Accident ◽  
Normal Operation ◽  
Water Radiolysis ◽  
Chain Reactions ◽  
Decay Heat ◽  
Severe Accidents ◽  
Heat Curve ◽  
Fukushima Daiichi

Although the water radiolysis, decomposition of water by radiation, is a well-known phenomenon the exact mechanism is not well characterized especially for severe accidents. The author first reviewed the water radiolysis phenomena in LWRs during normal operation to severe accidents (e.g., TMI- and Chernobyl accidents) and performed a scoping estimation of the amount of radiological hydrogen generation, accumulation and release for the Fukushima Daiichi accident. The estimation incorporates the decay heat curve after a reactor trip combined with G-values. As much as 450 cubic meters-STP of accumulated hydrogen gas is estimated to be located inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1. When a set of radiological chain reactions are incorporated the resultant reverse reactions substantially reduce the hydrogen generation, even when removal of molecular products (i.e., oxygen and hydrogen) is assumed stripped rapidly from boiling water through bubbles. Even in the most favorable configuration a typical amount of hydrogen gas reduces to only several tens of cubic meters. Finally, the author tested a new mechanism, “radiation-induced electrolysis,” which had been applied to his corrosion studies for last several years. His theory has been verified with the published in-pile test data, although he has never tried to apply this to his severe accident study. The predicted results indicated that the total inventory of hydrogen gas inside RPV may reach as much as 1000 cubic meters in just 3 hours during the SBO due to a high decay heat soon after the reactor trip through this process.

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