Analysis of Accident Progression of Fukushima Daiichi NPP by the SAMPSON Code

2014 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hiroaki Suzuki ◽  
Hideo Mizouchi ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Water Injection ◽  
Heat Removal ◽  
Fission Products ◽  
Severe Accident ◽  
Core Material ◽  
Decay Heat ◽  
The Core ◽  
Fukushima Daiichi ◽  
Fire Engine

This paper describes analysis results of the early phase accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Unit 1 by the severe accident analysis code SAMPSON. The isolation condensers were the only devices for decay heat removal at Unit 1, but they stopped after the loss of AC and DC powers. Since there were no decay heat removal for about 14 hours after their termination until the start of alternative water injection into the core by the fire engine, the core melt and the reactor pressure vessel (RPV) bottom failure occurred resulting in large amount of fission products release into the environment. The original SAMPSON was improved by adding new modellings for the phenomena which have been deemed specific to the Fukushima Daiichi NPP: (1) deterioration of SRV gaskets and (2) buckling of in-core-monitor housings which caused the early steam leakage from the core into the drywell, and (3) melt of the in-core-monitor housings in the lower plenum of the RPV. The analysis results showed that (1) 55.3% of UO2 of the initial loading and 66.1% of the core material including UO2, zircaloy, steel and control materials had melted down into the pedestal of the drywell, (2) the amount of Hydrogen generated by Zr-H2O reaction was 686 kg, (3) amount of Cs element released from fuels was 61 kg which was 72% of the total Cs element which was included in fuels at the initiation of the accident, and (4) 18.3% of the corium which fell into the pedestal was one large lump and the 81.7% was particulate corium.

Function of Isolation Condenser of Fukushima Unit-1 Nuclear Power Plant

2012 ◽  
Author(s):  
Masanori Naitoh ◽  
Hiroaki Suzuki ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Failure Time ◽  
Sea Water ◽  
Water Injection ◽  
Heat Removal ◽  
Severe Accident ◽  
Power Station ◽  
Fukushima Daiichi ◽  
Fire Engine

The Tohoku Region Pacific Coast Earthquake with magnitude 9.0 occurred at 2:46 PM of March 11th, 2011, followed by a huge Tsunami. The Fukushima Daiichi nuclear power station suffered serious damages from the Tsunami, involving core melt and release of large amount of fission products to an environment. The station blackout (SBO) occurred due to submergence of emergency equipment by the sea water. The isolation condenser (IC) was the only device for decay heat removal at the unit-1 of the Fukushima Daiichi nuclear power station after the reactor scram. The IC function was analyzed with a severe accident analysis code SAMPSON. The analysis results showed that (1) core melt resulting in RPV failure occurred since the IC operation was limited because it was not designed as a countermeasure to mitigate severe accident progression in Japan and (2) even assuming the continuous IC operation after the SBO to mitigate severe accident progression, the RPV failure occurred at 18:52, March 12th. However, since the alternate water injection by a fire engine was actually ready to start at 5:46, March 12th, which was earlier than calculated RPV failure time, the RPV failure could be prevented by continuous IC operation.

Development of High Efficiency Containment Venting System by Using AgX

2016 ◽  
Author(s):  
Tadashi Narabayashi ◽  
Yuuhei Sugano ◽  
Hiroki Imaeda ◽  
Go Chiba ◽  
Nobuaki Sato ◽  
...  
Keyword(s):  
High Efficiency ◽  
Water Injection ◽  
Heat Removal ◽  
Fission Products ◽  
Radioactive Material ◽  
Severe Accident ◽  
Test Facility ◽  
Silver Coating ◽  
Fukushima Daiichi ◽  
Filter Thickness

Fukushima Daiichi NPP accident would be terminated, if sufficient accident countermeasures, such as water proof door, mobile power, etc [1, 2]. In case of Europe, it had already installed the heat removal system and filtered containment venting system (FCVS) from the lessons of TMI and Chernobyl Accidents. The new regulatory standard in Japan, the filtered vent system (FCVS) should be installed, and prevent the radioactive material in case of the severe accident and the overpressure breakage prevention of a primary containment vessel (PCV) and also the robustization of the FCVS. The authors examined the severe accident process in the 2nd unit of Fukushima Daiichi NPS, and found the vent by FCVS should be done before water injection into the core. The PCV spray and water injection into the pedestal basement should be also the countermeasures to the severe accident. Countermeasures for an intentional aircraft collision should be installed too. Upon occurrence of a severe accident (SA), vent gas with radioactive fission products is blown out to a scrubbing pool through numerous venturi nozzles. Mist in steam moves upward to a metal fiber filter through a multi-hole baffle plate. After the mist is removed by that filter, radioactive methyl iodine (CH3I) is captured on the surface of a molecular sieve or AgX, made from zeolite particles with silver coating. A FCVS visualized test facility was installed at Hokkaido University. An AgX filter is used down-stream of the scrubbing pool and metal fiver filter. Thickness of AgX filter is very important parameter to obtain enough decontamination factor (DF). The DF for the radioactive iodine exceeds 10,000 at bed depth (AgX filter thickness) greater than 75mm.

Severe Accident Analysis With Spatial Discretized Model by MAAP: Part 1 — Parametric Study on Fukushima-Daiichi Unit-2

2018 ◽  
Author(s):  
Kenichi Kanda ◽  
Yoshihisa Nishi ◽  
Kazuma Abe ◽  
Satoshi Nishimura ◽  
Koichi Nakamura ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Water Injection ◽  
Fission Products ◽  
Severe Accident ◽  
Accident Analysis ◽  
Detailed Model ◽  
Parametric Studies ◽  
Fukushima Daiichi ◽  
Alternative Water ◽  
Small Parts

Accident analyses of the Fukushima-Daiichi unit-2 nuclear power plant were performed with MAAP (Modular Accident Analysis Program) version 5.03. We assumed RCIC, SRV operation and alternative water injection in order to reproduce the measured pressure and temperature values in RPV and PCV. From parametric studies, it was found that the analysis results were in good agreement with the measured data. In this paper, the results of the parametric studies are reported. Furthermore, spatial discretization of compartments (such as rooms in the reactor building, etc.) into small parts successfully demonstrated the transient distribution and deposition of fission products (FPs) across the rooms. Such special discretization is particularly important for the forensic investigation of severe accidents and the deposited amount in the R/B might be estimated by using this detailed model.

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.

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.

Severe accident simulation for VVER-1000 reactor using ASTEC-V2.1.1.3

Kerntechnik ◽  
2021 ◽  
Vol 86 (6) ◽  
pp. 454-469
Author(s):  
S. H. Abdel-Latif
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Fission Products ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Potential Core ◽  
Safety Requirements ◽  
The Core ◽  
Active Safety Systems ◽  
Reactor Pressure

Abstract The station black-out (SBO) is one of the main accident sequences to be considered in the field of severe accident research. To evaluate a nuclear power plant’s behavior in the context of this accident, the integral ASTEC-V2.1.1.3 code “Accident Source Term Evaluation Code” covers sequences of SBO accidents that may lead to a severe accident. The aim of this work is to discuss the modelling principles for the core melting and in-vessel melt relocation phenomena of the VVER-1000 reactor. The scenario of SBO is simulated by ASTEC code using its basic modules. Then, the simulation is performed again by the same code after adding and activating the modules; ISODOP, DOSE, CORIUM, and RCSMESH to simulate the ex-vessel melt. The results of the two simulations are compared. As a result of SBO, the active safety systems are not available and have not been able to perform their safety functions that maintain the safety requirements to ensure a secure operation of the nuclear power plant. As a result, the safety requirements will be violated causing the core to heat-up. Moreover potential core degradation will occur. The present study focuses on the reactor pressure vessel failure and relocation of corium into the containment. It also discusses the transfer of Fission Products (FPs) from the reactor to the containment, the time for core heat-up, hydrogen production and the amount of corium at the lower plenum reactor pressure vessel is determined.

scholarly journals FISSION-PRODUCT RELEASES FROM A PHWR TERMINAL DEBRIS BED

2016 ◽  
Vol 5 (1) ◽  
pp. 95-105 ◽  
Author(s):  
M.J. Brown ◽  
D.G. Bailey
Keyword(s):  
Fission Product ◽  
Heavy Water ◽  
Fission Products ◽  
Severe Accident ◽  
Water Reactor ◽  
Decay Heat ◽  
The Core ◽  
Heavy Water Reactor ◽  
Release Model ◽  
Tank Water

During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model.

Performance Analysis of a Novel PXS With Organic Rankine Cycle (ORC) System

2013 ◽  
Author(s):  
Shengjun Zhang
Keyword(s):  
Nuclear Power ◽  
Organic Rankine Cycle ◽  
Heat Removal ◽  
Rankine Cycle ◽  
Working Fluid ◽  
Heat Transfer Area ◽  
Decay Heat ◽  
The Core ◽  
Primary Fluid ◽  
Residual Heat

With the increasing of core thermal power of the nuclear power plant, the decay heat of the core increases in the accident. Therefore, the heat removal capacity of the PXS should be enhanced to fulfill the requirement of core safety. A new scheme is put forward to improve the cooling capacity of PXS and provides long-term power for station blackout (SBO) accident or loss of normal feedwater. In this system, the Organic Rankine Cycle is incorporated between the hot leg and cold leg of PRHR. The decay heat of the core is the heat source and the cooling pool outside the containment is the cool source. The natural circulation of the primary loop is established due to the density difference. The primary fluid flows into the evaporator of the ORC system, where the working fluid of the ORC system is evaporated. Then the temperature of the primary fluid is decreased. The vaporized working fluid drives the expander, which is coaxially fixed with the fluid pump, to generate the power. Finally, the exhausted vapor flows into the condenser and the residual heat is discharged outside of the containment simultaneously. The working fluid in the condenser is pumped into the evaporator by the fluid pump for liquid supplement and the cycle keeps on working continuously. A steady state analysis is performed on a 1700MWe nuclear power plant with ORC as the heat removal system. The heat transfer area of the ORC evaporator is fixed as 487.7m2, which is the same as the area of PRHR HX. The efficiencies of fluid pump and expander of ORC system are assumed as 0.75 and 0.8, respectively. The decay heat of the core is about 67.62MWe, which is 1.38% of the core full power. The working fluids are screened and R141b offers excellent performance. The efficiency of fluid pump and expander are assumed as 0.75 and 0.8, respectively. The condensing temperature is assumed as 80°C and the evaporating temperature is 160°C. The results show that 7.83MWe will be generated by the ORC system and the heat transfer area of the condenser is about 994.5m2. The residual heat of 59.79MWe will be discharged to the water tank outside the containment.

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