Evaluation of a Severe Accident Management Strategy for Boiling Water Reactors—Drywell Flooding

1994 ◽  
Vol 106 (2) ◽  
pp. 139-154 ◽  
Author(s):  
Donghan Yu ◽  
Leiming Xing ◽  
William E. Kastenberg ◽  
David Okrent
Keyword(s):  
Management Strategy ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactors ◽  
Accident Management ◽  
Water Reactors

Passive depressurization accident management strategy for boiling water reactors

2015 ◽  
Vol 284 ◽  
pp. 176-184 ◽  
Author(s):  
Maolong Liu ◽  
Nejdet Erkan ◽  
Yuki Ishiwatari ◽  
Koji Okamoto
Keyword(s):  
Management Strategy ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Accident Management ◽  
Water Reactors

Behavior of Simulated Molten Metal at Channel Modeled Boiling Water Reactors

2017 ◽  
Author(s):  
Yutaro Hihara ◽  
Kota Matsuura ◽  
Hideaki Monji ◽  
Yutaka Abe ◽  
Akiko Kaneko ◽  
...  
Keyword(s):  
Experimental Data ◽  
Numerical Simulation ◽  
Numerical Analysis ◽  
Molten Metal ◽  
Numerical Results ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactors ◽  
Simulation Code ◽  
Water Reactors

When a severe accident occurs, decommissioning work becomes important task. In the decommissioning work after the severe accident, establishing the way to estimate the sedimentation place of molten debris is important. However, the technique to estimate exactly sedimentation place has not been enough. Therefore, the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior is under development. The comparison between experimental and numerical results is necessary to clarify the validity of the numerical analysis code. This study provides the experimental data for a BWR to examine the numerical simulation code in order to contribute to progress of the decommissioning work.

Features of a control blade degradation observed in situ during severe accident conditions in boiling water reactors

2019 ◽  
Vol 56 (5) ◽  
pp. 440-453 ◽  
Author(s):  
Anton Pshenichnikov ◽  
Saishun Yamazaki ◽  
David Bottomley ◽  
Yuji Nagae ◽  
Masaki Kurata
Keyword(s):  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactors ◽  
Water Reactors ◽  
Accident Conditions

Retention of Molten Corium in Calandria Vessel of PHWR With Moderator Drain Pipe

2020 ◽  
Vol 7 (1) ◽  
Author(s):  
Pradeep Pandey ◽  
Parimal P. Kulkarni ◽  
Arun Nayak ◽  
Sumit V. Prasad
Keyword(s):  
Thermal Load ◽  
Management Strategy ◽  
Severe Accident ◽  
Drain Pipe ◽  
Severe Accidents ◽  
External Cooling ◽  
Accident Management ◽  
Water Reactors ◽  
Containment Integrity

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.

scholarly journals Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code

2019 ◽  
Vol 2019 ◽  
pp. 1-18
Author(s):  
Sergey Galushin ◽  
Pavel Kudinov
Keyword(s):  
Severe Accident ◽  
Boiling Water Reactors ◽  
Light Water Reactors ◽  
Significant Discrepancy ◽  
The Core ◽  
Accident Management ◽  
Core Components ◽  
Water Reactors ◽  
Delayed Time ◽  
Containment Integrity

Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.

Recirculation Pump Speed Control System Upgrades in Nuclear Boiling Water Reactors

2008 ◽  
Author(s):  
James W. Morgan
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Control ◽  
Nuclear Power ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Pump Speed ◽  
Water Reactors ◽  
Generator Sets

The nuclear power industry is faced with determining what to do with equipment and instrumentation reaching obsolescence and selecting the appropriate approach for upgrading the affected equipment. One of the systems in a nuclear power plant that has been a source of poor reliability in terms of replacement parts and control performance is the reactor recirculation pump speed/ flow control system for boiling water reactors (BWR). All of the operating BWR-3 and BWR-4’s use motor-generator sets, with a fluid coupled speed changer, to control the speed of the recirculation water pumps over the entire speed range of the pumps. These systems historically have had high maintenance costs, relative low efficiency, and relatively inaccurate speed control creating unwanted unit de-rates. BWR-5 and BWR-6 recirculation flow control schemes, which use flow control valves in conjunction with two-speed pumps, are also subject to upgrades for improved performance and reliability. These systems can be improved by installing solid-state adjustable speed drives (ASD), also known as variable frequency drives (VFD), in place of the motor-generator sets and the flow control valves. Several system configurations and ASD designs have been considered for optimal reliability and return on investment. This paper will discuss a highly reliable system and ASD design that is being developed for nuclear power plant reactor recirculation water pump controls. Design considerations discussed include ASD topology, controls architecture, accident, transient and hydraulic analyses, potential reactor internals modifications, installation, demolition and economic benefits.

The Steam Turbine Retrofitting of Six Boiling Water Reactor Units Using Innovative Engineering and Laser Scanning Techniques

2009 ◽  
Author(s):  
Mike Jones ◽  
David J. Nelmes
Keyword(s):  
Steam Turbine ◽  
Laser Scanning ◽  
Laser Scanner ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Extensive Experience ◽  
Turbine Generators ◽  
The Usa ◽  
Water Reactors ◽  
Scan Data

Alstom Power is executing the steam turbine retrofit of six nuclear units for Exelon Generation in the USA. The existing turbine-generators are an 1800 RPM General Electric design originally rated at 912 MWe and 1098 MWe and powered by Boiling Water Reactors. 18 Low Pressure inner modules will be replaced, with the first due to be installed in March 2010. This project is particularly challenging — the aggressive retrofit installation schedule is compounded by the requirement to handle radioactively contaminated equipment and also comply with demanding regulations applicable to BWR plant. The author’s company has extensive experience in the steam turbine retrofit business, having supplied around 800 retrofit cylinders globally since the 1970’s. However, this LP upgrade challenges the established techniques used in the business and requires extraordinary effort. Traditional retrofit engineering and installation principles have been interrogated and developed to meet the specific requirements of this project. Innovative techniques are introduced, including the extensive use of the Leica HDS 6000 laser scanner to model the existing plant. The approach has advanced the field of steam turbine retrofit design and installation significantly. The first section of this paper focuses on the extraordinary considerations of the project and the challenges surrounding BWR plant. The second part describes the laser scanning technique and the application of scan data. It outlines the innovative solutions which have been developed.

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