scholarly journals Structural Integrity Evaluation of a Reactor Cavity during a Steam Explosion for External Reactor Vessel Cooling

Energies ◽  
2021 ◽  
Vol 14 (12) ◽  
pp. 3605
Author(s):  
Sang-Hyun Park ◽  
Kwang-Hyun Bang ◽  
Jong-Rae Cho
Keyword(s):  
Steam Explosion ◽  
Nuclear Power ◽  
Transient Analysis ◽  
Structural Integrity ◽  
Cooling System ◽  
Reactor Core ◽  
Ductile Failure ◽  
Nuclear Accidents ◽  
Great Damage ◽  
Severe Nuclear Accident

Nuclear power is a major source of electricity in the international community. However, a significant problem with nuclear power is that, if a severe nuclear accident occurs, radiation may leak and cause great damage. As such, research on nuclear safety has become increasingly popular worldwide. In this paper, the structural integrity of a reactor cavity during a steam explosion—one kind of the aforementioned severe nuclear accidents—was evaluated. Steam explosions are primarily caused by fuel–coolant interactions (FCI), and result from issues in the cooling system that discharges the melt from the reactor core to the outside. A steam explosion can damage the nuclear power plant, and radiation leakage, the greatest concern, may occur. In the Chernobyl or Fukushima Daiichi accidents, significant radiation leakages resulted in damages extending beyond the country of origin. In this paper, a steam explosion was simulated using values given by the transient analysis code for explosive reactions (TRACER-II)—the only steam explosion code in Korea. The walls of the reactor cavity were modeled after the APR-1400 currently operating in Korea. The integrity of the concrete, rebars, and liner plate in the reactor cavity during a steam explosion was evaluated in terms of stress and ductile failure strain limits.

Precautions at Fukushima That Would Have Suppressed the Accident Severity

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Kenji Iino ◽  
Ritsuo Yoshioka ◽  
Masao Fuchigami ◽  
Masayuki Nakao
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Water Injection ◽  
Reactor Core ◽  
Radioactive Material ◽  
Injection System ◽  
Tsunami Waves ◽  
Ac Power ◽  
Pressure Water ◽  
High Pressure Coolant

Abstract The Great East Japan Earthquake on Mar. 11, 2011 triggered huge tsunami waves that attacked Fukushima Daiichi Nuclear Power Plant (Fukushima-1). Units 1, 3, and 4 had hydrogen explosions. Units 1–3 had core meltdowns and released a large amount of radioactive material. Published investigation reports did not explain how the severity of the accident could have been prevented. We formed a study group to find: (A) Was the earthquake-induced huge tsunami predictable at Fukushima-1? (B) If it was predictable, what preparations at Fukushima-1 could have avoided the severity of the accident? Our conclusions were: (a) The tsunami that hit Fukushima-1 was predictable, and (b) the severity could have been avoided if the plant had prepared a set of equipment, and most of all, had exercised actions to take against such tsunami. Necessary preparation included: (1) a number of direct current (DC) batteries, (2) portable underwater pumps, (3) portable alternating current (AC) generators with sufficient gasoline supply, (4) high voltage AC power trucks, and (5) drills against extended loss of all electric power and seawater pumps. This set applied only to this specific accident. A thorough preparation would have added (6) portable compressors, (7) watertight modification to reactor core isolation cooling system (RCIC) and high pressure coolant injection system (HPCI) control and instrumentation, and (8) fire engines for alternate low pressure water injection. Item (5), i.e., to study plans and carry out exercises against the tsunami would have identified all other necessary preparations.

Self-Propelling Cooling Systems: Back-Fitting Passive Cooling Functions to Existing Nuclear Power Plants

2012 ◽  
Author(s):  
Dominik von Lavante ◽  
Dietmar Kuhn ◽  
Ernst von Lavante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Electrical Power ◽  
Reactor Core ◽  
Passive Cooling ◽  
Spent Fuel ◽  
Cooling Systems ◽  
Spent Fuel Pool

The present paper describes a back-fit solution proposed by RWE Technology GmbH for adding passive cooling functions to existing nuclear power plants. The Fukushima accidents have high-lighted the need for managing station black-out events and coping with the complete loss of the ultimate heat sink for long time durations, combined with the unavailability of adequate off-site supplies and adequate emergency personnel for days. In an ideal world, a nuclear power plant should be able to sustain its essential cooling functions, i.e. preventing degradation of core and spent fuel pool inventories, following a reactor trip in complete autarchy for a nearly indefinite amount of time. RWE Technology is currently investigating a back-fit solution involving “self-propelling” cooling systems that deliver exactly this long term autarchy. The cooling system utilizes the temperature difference between the hotter reactor core or spent fuel pond with the surrounding ultimate heat sink (ambient air) to drive its coolant like a classical heat machine. The cooling loop itself is the heat machine, but its sole purpose is to merely achieve sufficient thermal efficiency to drive itself and to establish convective cooling (∼2% thermal efficiency). This is realized by the use of a Joule/Brayton Cycle employing supercritical CO2. The special properties of supercritical CO2 are essential for this system to be practicable. Above a temperature of 30.97°C and a pressure of 73.7bar CO2 becomes a super dense gas with densities similar to that of a typical liquid (∼400kg/m3), viscosities similar tothat of a gas (∼3×105Pas) and gas like compressibility. This allows for an extremely compact cooling system that can drive itself on very small temperature differences. The presented parametric studies show that a back-fitable system for long-term spent fuel pool cooling is viable to deliver excess electrical power for emergency systems of approximately 100kW. In temperate climates with peak air temperatures of up to 35°C, the system can power itself and its air coolers at spent fuel pool temperatures of 85°C, although with little excess electrical power left. Different back-fit strategies for PWR and BWR reactor core decay heat removal are discussed and the size of piping, heat exchangers and turbo-machinery are briefly evaluated. It was found that depending on the strategy, a cooling system capable of removing all decay heat from a reactor core would employ piping diameters between 100–150mm and the investigated compact and sealed turbine-alternator-compressor unit would be sufficiently small to be integrated into the piping.

Inspection and Evaluation Guidelines for Light Water Reactor Core Internals in Japan

2002 ◽  
Author(s):  
Kimihito Takeuchi ◽  
Naoto Iizuka ◽  
Masashi Kameyama ◽  
Haruo Fujimori ◽  
Yuichi Motora ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Structural Integrity ◽  
Reactor Core ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Integrity Assessment ◽  
Engineering Society ◽  
Safety Function ◽  
Internal Component

There have been many cracking experiences of light water reactor (LWR) core internals worldwide in the past. Thermal and Nuclear Power Engineering Society in Japan (TENPES) has organized a committee to prepare technically reasonable and appropriate inspection and evaluation guidelines (I&E guidelines) for core internals. This committee consists of scholars and representatives from electric utilities and nuclear plant vendors in Japan. I&E guidelines, which cover a rational inspection plan, structural integrity assessment and repair methods, have been developed considering nuclear safety function and structural strength of each core internal component. For BWR reactors, the development of I&E guidelines cover major core internal components like shroud support, core shroud, top guide, core plate, ICM and CRD housing, core spray piping and sparger, jet pump etc. For PWR reactors, the development of I&E guidelines cover baffle former bolts, barrel former bolts, core barrel weld, bottom mounted instrumentation, etc. The I&E guidelines will be completed by the end of March 2002. The basic concept of the guidelines, and a guideline for shroud support of a BWR as an example, are shown in this paper.

Optimizing Maintenance Strategy of a Reactor Pressure Vessel Using 3D-CFD and FEM Based Probabilistic Pressurized Thermal Shock Analysis

2017 ◽  
Author(s):  
Xiaoyong Ruan ◽  
Toshiki Nakasuji ◽  
Kazunori Morishita
Keyword(s):  
Spatial Distribution ◽  
Thermal Shock ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Cooling System ◽  
Reactor Pressure Vessel ◽  
Pressurized Thermal Shock ◽  
Core Cooling ◽  
Reactor Pressure

The structural integrity of a reactor pressure vessel (RPV) is important for the safety of a nuclear power plant. When the emergency core cooling system (ECCS) is operated and the coolant water is injected into the RPV due to a loss-of-coolant accident (LOCA), the pressurized thermal shock (PTS) loading takes place. With the neutron irradiation, PTS loading may lead a RPV to fracture. Therefore, it is necessary to evaluate the performance of RPV during PTS loading to keep the reactor safety. In the present study, optimization of RPV maintenance is considered, where two different attempts are made to investigate the RPV integrity during PTS loading by employing the deterministic and probabilistic methodologies. For the deterministic integrity evaluation, 3D-CFD and finite element method (FEM) simulations are performed, and stress intensity factors (SIFs) are obtained as a function of crack position inside the RPV. As to the probabilistic integrity evaluation, on the other hand, a more accurate spatial distribution of SIF on the RPV is calculated. By comparing the distribution thus obtained with the fracture toughness included as a part of the master curve, the dependence of fracture probabilities on the position inside the RPV is obtained. Using the spatial distribution of fracture probabilities in RPV, the priority of the inspection and maintenance is finally discussed.

scholarly journals Early quantification of reliability for a safety critical and control system: a case study of reactor core cooling system

2018 ◽  
Vol 7 (2.12) ◽  
pp. 248
Author(s):  
Vinay Kumar ◽  
Suraj Gupta ◽  
Anil Kumar Tripathi
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Safety Critical ◽  
Core Cooling ◽  
And Control

Using Probabilistic Reliability analysis for Quantifying reliability of a system is already a common practice in Reliability Engineering community. This method plays an important role in analyzing reliability of nuclear plants and its various components. In Nuclear Power Plants Reactor Core Cooling System is a component of prime importance as its breakdown can disrupt Cooling System of power plant. In this paper, we present a framework for early quantification of Reliability and illustrated with a Safety Critical and Control System as case study which runs in a Nuclear Power Plant.  

Modification to Shield Tank Overpressure Protection in a CANDU Reactor for Beyond Design Basis Event

2019 ◽  
Author(s):  
Michael Huang ◽  
Khurram Khan ◽  
Ali Etedali-Zadeh ◽  
Jefferson Tse ◽  
Bing Li
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Lower Probability ◽  
Damage State ◽  
Core Damage ◽  
Design Basis ◽  
System Pressure ◽  
Severe Event ◽  
Large Heat

Abstract The Shield Tank and End Shield Cooling System in the CANDU reactor contains a large volume of light water surrounding the Calandria and circulates water to remove heat that arises from the reactor core and Moderator. In a beyond design basis event that results in a severe event, progression in the absence of mitigating cooling actions could result in a large heat load being transferred to the water inside the shield tank from the calandria wall causing shield tank failure due to over pressurization. Following the 2011 events at Fukushima Daiichi Nuclear Power Plant, the adequacy of system pressure relief was assessed against severe events. Emergency mitigating equipment tie-ins for water make-up will likely limit the core damage state and prevent the need to protect the shield tank. However, Shield Tank Overpressure Protection (STOP) has been installed against severe event conditions pursuant to the CANDU defense-in-depth safety philosophy. A larger open vent line has been installed at some CANDU units on the top of the shield tank outside containment. This design routes the vent piping high enough to preclude any venting under any operational configuration and discharges back into the containment through an existing spare penetration. Vent piping is designed as Nuclear Class 2 in accordance with ASME BPVC Section III. Assessment of stresses in the modification piping was also completed for BDBEs including for a lower probability seismic event, steam venting and corresponding higher pressure and temperature conditions.

scholarly journals Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit

2008 ◽  
Vol 2008 ◽  
pp. 1-8
Author(s):  
A. Kaliatka ◽  
E. Uspuras ◽  
M. Vaisnoras
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Thermal Inertia ◽  
Severe Accident ◽  
Cooling Circuit ◽  
High Thermal Inertia ◽  
The Moment ◽  
Control Rods

The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was shut down on December 31, 2004, and Unit 2 is to be operated until the end of 2009. Despite of this fact, severe accident management guidelines for RBMK-1500 reactor at Ignalina NPP are prepared. In case of beyond design basis accidents, it can occur that no water sources are available at the moment for heat removal from fuel channels. Specificity of RBMK reactor is such that the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. Therefore, the heat removal from RBMK-1500 reactor core using circuit for cooling of rods in control and protection system can be used as nonregular mean for reactor cooldown in case of BDBA. The heat from fuel channels, where heat is generated, through graphite bricks is transferred in radial direction to cooled CPS channels. This article presents the analysis of possibility to remove heat from reactor core in case of large LOCA by employing CPS channels cooling circuit. The analysis was performed for Ignalina NPP with RBMK-1500 reactor using RELAP5-3D and RELAP5 codes. Results of the analysis have shown that, in spite of high thermal inertia of graphite, this heat removal from CPS channels allows to slow down effectively the core heat-up process.

Analysis of Accident Progression of Fukushima Daiichi NPPs With SAMPSON Code

2013 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Severe Accident ◽  
Injection System ◽  
Part Load ◽  
Calculation Results ◽  
Fukushima Daiichi ◽  
Measured Pressure

The Fukushima Daiichi Nuclear Power Plant units 1, 2, and 3 had serious damages due to the huge earthquake and tsunami which occurred on March 11th 2011. Pressure transients in the reactor pressure vessels (RPVs) of the units 1, 2, and 3 were analyzed with the severe accident analysis code, SAMPSON for a few days from the scram until occurrence of depressurization. Since preliminary analysis results with the original SAMPSON showed difference from the measured data, the following phenomena were newly considered in the current analyses. For unit 1: Damage of a source range monitor, which is one of in-core monitors. For unit 2: Part load operation of the reactor core isolation cooling system. For unit 3: Part load operation of the high pressure coolant injection system. The calculation results showed fairly good agreements with the measured pressure data and showed RPV bottom damage for all the units resulting in falling of debris in the core region into the pedestal of the drywell.

Application of Thermal Hydraulic and Severe Accident Code SOCRAT/V3 to Bottom Water Reflood Experiment QUENCH-LOCA-1

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Alexander Vasiliev ◽  
Juri Stuckert
Keyword(s):  
Nuclear Power ◽  
Hydrogen Generation ◽  
Cooling System ◽  
Reactor Core ◽  
Severe Accident ◽  
Loss Of Coolant Accident ◽  
Post Test ◽  
Core Cooling ◽  
Modeling Code ◽  
Code Quality

This study aims to (1) use the thermal hydraulic and severe fuel damage (SFD) best-estimate computer modeling code SOCRAT/V3 for post-test calculation of QUENCH-LOCA-1 experiment and (2) estimate the SOCRAT code quality of modeling. The new QUENCH-LOCA bundle tests with different cladding materials will simulate a representative scenario for a loss-of-coolant-accident (LOCA) nuclear power plant (NPP) accident sequence in which the overheated (up to 1050°C) reactor core would be reflooded from the bottom by the emergency core cooling system (ECCS). The test QUENCH-LOCA-1 was successfully performed at the KIT, Karlsruhe, Germany, on February 2, 2012, and was the first test for this series after the commissioning test QUENCH-LOCA-0 conducted earlier. The SOCRAT/V3-calculated results describing thermal hydraulic, hydrogen generation, and thermomechanical behavior including rods ballooning and burst are in reasonable agreement with the experimental data. The results demonstrate the SOCRAT code’s ability for realistic calculation of complicated LOCA scenarios.

scholarly journals Evaluation of Pipe Whip Impacts on Neighboring Piping and Walls of the Ignalina Nuclear Power Plant

2002 ◽  
Author(s):  
Gintautas Dundulis ◽  
Ronald F. Kulak ◽  
Algirdas Marchertas ◽  
Evaldas Narvydas ◽  
Mark C. Petry ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Transient Analysis ◽  
Cooling System ◽  
Nonlinear Material ◽  
Reactor Safety ◽  
Concrete Wall ◽  
Concrete Walls ◽  
Reinforced Concrete Wall ◽  
Core Cooling ◽  
Group Distribution

Presented in this paper is the transient analysis of a Group Distribution Header (GDH) following a guillotine break at the end of the header. The GDH is the most important component of reactor safety in case of accidents. Emergency Core Cooling System (ECCS) piping is connected to the GDH piping such that, during an accident, coolant passes from the GDH into the ECCS. The GDH that is propelled into motion after a guillotine break can impact neighboring GDH pipes or the nearest wall of the compartment. Therefore, two cases are investigated: • GDH impact on an adjacent GDH and its attached piping; • GDH impact on an adjacent reinforced concrete wall. A whipping RBMK-1500 GDH along with neighboring concrete walls and pipelines is modeled using finite elements. The finite element code NEPTUNE used in this study enables a dynamic pipe whip structural analysis that accommodates large displacements and nonlinear material characteristics. The results of the study indicate that a whipping GDH pipe would not significantly damage adjacent walls or piping and would not result in a propagation of pipe failures.

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