scholarly journals MELCOR Modeling of Combined Accident Tolerant Fuel and Reactor Core Isolation Cooling System Operation

2020 ◽  
Author(s):  
Chris Faucett ◽  
Bradley Beeny ◽  
Karen Vierow Kirkland
Keyword(s):  
Hydrogen Production ◽  
Cooling System ◽  
Reactor Core ◽  
The United States ◽  
New Physics ◽  
Department Of Energy ◽  
Severe Accident ◽  
United States Government ◽  
Accident Analysis ◽  
Combined Use

Abstract The work presented in this paper presents new techniques for modeling the combined use of the Reactor Core Isolation Cooling (RCIC) System and Accident Tolerant Fuel (ATF) in a Boiling Water Reactor (BWR). With guidance from Sandia National Laboratories’ Severe Accident Analysis department, a MELCOR BWR model was developed from open source literature. The demonstration shown herein simulates BWR long-term station blackout (LTSBO) conditions with the Nuclear Regulatory Commission’s (NRC) MELCOR severe accident analysis code. By combining state-of-the-art MELCOR modeling practices with new, physics-based RCIC System and ATF MELCOR inputs, this BWR model provides a contemporary platform for BWR severe accident simulations. The authors are investigating the combined use of the RCIC System and ATF as a means of passively enhancing reactor safety. The benefits of this approach were evaluated by performing simulations using traditional fuel designs (i.e. Zircaloy cladding) and ATF with an iron-chromium-aluminum (FeCrAl) clad under BWR LTSBO conditions. ATF performance was evaluated using severe accident metrics, specifically event sequence timings and the hydrogen production rate from cladding oxidation. Preliminary results show delayed core degradation timelines with less hydrogen production for ATF simulations. Although the results are limited in scope, the presented analysis could easily be expanded to a full-scale uncertainty study that considers a range of severe accident boundary conditions. This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the U.S. Department of Energy or the United States Government.

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.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

Cryopreservation of human spermatozoa. IV. The effects of cooling rate and warming rate on the maintenance of motility, plasma membrane integrity, and mitochondrial function**Supported by grants from the National Institutes of Health, Bethesda, Maryland (R01-HD25949 and K04-HD00980); Parke-Davis; Methodist Hospital; and by the Office of Health and Environmental Research, United States Department of Energy, Washington, D.C., under contract DE-AC05-840R21400 with Martin Marietta Energy Systems.††Presented in part at the Annual Meeting of the American Society of Andrology, New Orleans, Louisiana, April 13 to 16, 1989, and at the Annual Meeting of the American Society of Andrology, Tampa, Florida, April 17 to 19, 1993.‡‡The submitted manuscript has been authored by a contractor of the United States Government under contract DE-AC05-840R21400. Accordingly, the United States Government retains a nonexclusive, royalty-free license to publish or reproduce the published form of this contribution, or allow others to do so, for U.S. Government purposes.

1993 ◽  
Vol 60 (5) ◽  
pp. 911-918 ◽  
Author(s):  
Mike A. Henry ◽  
Esther E. Noiles ◽  
Dayong Gao ◽  
Peter Mazur ◽  
John K. Critser
Keyword(s):  
United States ◽  
Annual Meeting ◽  
Membrane Integrity ◽  
The United States ◽  
Department Of Energy ◽  
Environmental Research ◽  
United States Government ◽  
United States Department ◽  
States Government ◽  
American Society

Hydrogen Production and Utilization in Petroleum Refineries: A Study of the U.S. Oil and Gas Industry

2011 ◽  
Author(s):  
Mohamed A. Mohamed ◽  
Radwa Soelem ◽  
Fares Attar ◽  
Nesrin Ozalp
Keyword(s):  
Hydrogen Production ◽  
Oil And Gas ◽  
Oil And Gas Industry ◽  
The United States ◽  
Department Of Energy ◽  
Oil Refining ◽  
Efficient Production ◽  
Production Plant ◽  
Gas Industry ◽  
The U.S

Petroleum refining industry in the United States is the largest in the world operating 148 refineries. These refineries contribute a major economic value to the U.S. market for providing the chemical industry with vital products. The economic gain, however, is challenged by the increasing competitiveness within the refining sector as well as the unpredictable oil prices. Furthermore, environmental obligations also have been recently advocating low emission rates that may entail additional operating costs to refineries. In this study, we analyze hydrogen production and utilization in the U.S. oil and gas industry to characterize its key role and trends in this energy-intensive industry. We referred to U.S. Department of Energy data and statistics of hydrogen production rates as well as we considered other elementary factors of refineries productivity such as; economics of crude oil, power consumption and chemical outputs. Considering the fact that hydrogen-dependent processes in refining count as a key element in oil refining; it is certainly that efficient production and implementation of hydrogen in processes such as hydro-cracking and hydro-desulfurization will result in cost saving opportunities for refineries. From this point of view, we highlight the economic and environmental advantages of solar cracking of natural gas as an alternative way of hydrogen production. Hydrogen production in refineries could possibly benefit from utilizing this alternative method on both local and global levels. Economically, this study explains how solar cracking could save about $62 million in hydrogen production for U.S. refineries. Even though the momentum of desulfurization acts are not yet strong in the U.S., major European refining investments are in jeopardy if not soon to utilize enhanced desulfurization facilities in response to demands of lower sulfur content of refined products. A comprehensive expenditures model is presented in this study to monitor primary areas of saving in hydrogen production from the early stages of establishing a hydrogen production plant. Further alternatives showing potential are also included as future considerations for the refinery sector.

Analysis of 3-Dimensional Hydrogen Behavior With Passive Containment Cooling

2017 ◽  
Author(s):  
Xingguan Huang ◽  
Tingzao Fu ◽  
Gaofeng Huang
Keyword(s):  
Cooling System ◽  
Water Film ◽  
Severe Accident ◽  
Accident Analysis ◽  
3 Dimensional ◽  
Film Evaporation ◽  
Lumped Parameter ◽  
Development Procedure ◽  
Hydrogen Analysis ◽  
Hydrogen Safety

Hydrogen safety is one of the most important issues for severe accident analysis. Comparing with the lumped parameter code, 3-D CFD code GASFLOW has more advantages for analyzing hydrogen related phenomena [1]. However, due to the lacking of passive containment cooling system (PCS) model, GASFLOW is inapplicable for hydrogen analysis of PWR with PCS. This paper shows the development procedure of PCS model, including the model introduction and validation. The main functions of the PCS model are the calculation of falling water film thickness, heat transfer between wall and film, and film evaporation. Then, the hydrogen safety analysis by GASFLOW with PCS model is performed for CAP1400.

Evaluating the fracture toughness of reactor pressure vessel (RPV) materials subject to embrittlement**Some portions of this chapter have been gleaned from Chapter 3 of: International Atomic Energy Agency, Integrity of Reactor Pressure Vessels in Nuclear Power Plants: Assessment of Irradiation Embrittlement Effects in Reactor Pressure Vessel Steels, IAEA Nuclear Energy Series NP-T-3.11, IAEA, Vienna (2009), a chapter authored by the first author of this chapter (no attribution in the IAEA document)Notice: This manuscript has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the US Department of Energy. The United States Government retains and the publisher by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

2015 ◽  
pp. 295-332
Author(s):  
R.K. Nanstad ◽  
W.L. Server ◽  
M.A. Sokolov ◽  
M. Brumovský
Keyword(s):  
United States ◽  
Pressure Vessel ◽  
Power Plants ◽  
The United States ◽  
Reactor Pressure Vessel ◽  
Department Of Energy ◽  
United States Government ◽  
Irradiation Embrittlement ◽  
States Government ◽  
Reactor Pressure

The Development of the Evolutionary BWR

2006 ◽  
Author(s):  
A. Murase ◽  
M. Nakamaru ◽  
M. Kuroki ◽  
Y. Kojima ◽  
S. Yokoyama
Keyword(s):  
Cooling System ◽  
Reactor Core ◽  
Heat Removal ◽  
Safety Performance ◽  
Severe Accident ◽  
Main Role ◽  
Isolation System ◽  
Near Term ◽  
Core Cooling ◽  
Control Rods

Considering the delay of the fast breeding reactor (FBR) development, it is expected that the light water reactor will still play the main role of the electric power generation in the 2030’s. Accordingly, Toshiba has been developing a new conceptual ABWR as the near-term BWR. We tentatively call it AB1600. The AB1600 has introduced the hybrid active/passive safety system in order to improve countermeasure against severe accident (SA). At the same time, we have made the simplification of the overall plant systems in order to improve economy. The simplification of the AB1600 is based on the proven technologies. To retain the safety performance superior or equivalent to the current ABWR and to strengthen the countermeasure against SA, the AB1600 has introduced the passive systems such as the passive containment cooling system (PCCS), the gravity driven core cooling system (GDCS) and the isolation condenser (IC). While we retain the safety performance superior or equivalent to the current ABWR, we have made the simplification of the safety systems. We could eliminate the high pressure core flooder system (HPCF) and the reactor core isolation system (RCIC) by extending the height of reactor pressure vessel (RPV) two meters. To achieve simplification of reactor systems, we have reduced the number of fuel bundles and the number of control rods by adopting large bundle that has a bundle pitch 1.2 times wider than that of the current ABWR. In the 1600MWe class, the number of fuel bundles could be reduced to 600 from 872 of the current ABWR, and the number of control rods could be reduced to 137 from 205 of the current ABWR. Because the reactor internal pump (RIP) of the current ABWR has sufficient performance capacity and the improvement of fuel characteristics from the current fuel enables the operation at lower core flow, the number of RIPs could be decreased from ten to eight. Furthermore, we have reduced the number of divisions of emergency core cooling system (ECCS)/heat removal system to two from three of the current ABWR. This configuration change contributes to reduce the amount of resources of not only reactor systems but also auxiliary systems. In the previous paper, the AB1600 had four low pressure flooder systems (LPFLs). We have studied about the possibility of reduction of LPFLs to two from four by providing the LPFL with alternative injection lines. This change is expected to contribute to reduce the total number of ECCS pumps and the capacity of emergency AC power.

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.

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