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MAAP - Modular Accident Analysis Program

Modular Accident Analysis Program (MAAP) Version 4 (EPRI owned and licensed computer software)

BackgroundCorium Pool Modeling in the Lower Plenum and MAAP Services

The Modular Accident Analysis Program (MAAP) Version 4 - an Electric Power Research Institute (EPRI) owned and licensed computer software - is a fast-running computer code that simulates the response of light water and heavy water moderated nuclear power plants for both current and Advanced Light Water Reactor (ALWR) designs.  It can simulate Loss-Of-Coolant Accident (LOCA) and non-LOCA transients for Probabilistic Risk Analysis (PRA) applications as well as severe accident sequences, including actions taken as part of the Severe Accident Management Guidelines (SAMGs).  There are several parallel versions of MAAP4 for BWRs, PWRs, CANDU designs, FUGEN design and the Russian VVER PWR design. 

Originally developed by Fauske & Associates, LLC (FAI) as part of the Industry Degraded Core Rulemaking (IDCOR) program, FAI has developed and maintained the code under the sponsorship of (EPRI) and the MAAP Users Group (MUG).  MAAP4 and its predecessor, MAAP3B, have been used by the nuclear industry throughout the world for more than two decades as an engineering tool to support PRA and severe accident analysis.

MAAP4 FEATURES

The MAAP4 input file processor allows users the capability to simulate operator actions for LOCA and non-LOCA transients, thus allowing the code to simulate the full spectrum of plant response to all types of accidents.  Several utilities have even successfully modeled their entire Emergency Operating Procedures (EOPs) using MAAP4.  MAAP4 also includes a graphical interface, MAAP4-GRAAPH, to increase the analytical capability as well as enhance the review function.  Specifically, the user can interactively interface with the code during execution to modify the status of on-site power, pumps, valves, etc., as well as to directly observe the results.

MAAP4 BENEFITS

The MAAP4 code builds on the substantial modeling enhancements over MAAP3B to assess the approach to, and progression of, severe accidents in BWRs and PWRs operated in the western hemisphere, and extends these to the plant-specific features that are presently represented in the state-of-the-art models for PRAs/PSAs, SAMGs, EOPs, and full scope control room simulators.  MAAP4 is a fast-running computer code capable of modeling an array of accident sequences in a short period of time, thus providing plant PRA staff a powerful tool to answer day to day issues that come up in the operation of a nuclear power plant.

MAAP4 EXPERIENCE/SOLUTIONS

The MAAP4 code has been used to perform severe accident analysis and associated severe accident phenomena, including hydrogen generation and combustion, direct-containment heating, rapid pressurization due to steaming, core concrete interactions, fission product releases, transport and deposition, etc.  The MAAP4 code is also used extensively in the probabilistic risk/safety assessment (PRA/PSA) arena as well for success criteria evaluations, human reliability analyses (HRA) and Level II source term evaluations, etc.

MODULAR ACCIDENT ANALYSIS PROGRAM (MAAP) – VERSION 5

(EPRI owned and licensed computer software)

BackgroundMAAP Services and PWR RCS Coolant Loop View

Following the accident at Three Mile Island Unit 2, the nuclear power industry developed the MAAP (EPRI owned and licensed computer software) computer code as part of the industry degraded core rulemaking (IDCOR) program. Its objective was to provide a useful tool for analyzing the consequences of a wide range of postulated plant transients and severe accidents for current plant designs and Advanced Light Water Reactors (ALWRs). The code can predict the progression of accident scenarios to a safe, stable, coolable state within the core or it can predict the occurrence of vessel failure and model the containment performance with successful debris cooling or pressurization of containment to a pre-defined failure condition. MAAP5 is the latest version of the suite of MAAP computer codes designed specifically to perform accident and severe analyses for numerous nuclear plant designs.

Additional work to augment the code to include Mark 1 containment and BWR-specific models after the Fukushima continues.

MAAP5 SEVERE ACCIDENT SIMULATOR AND TRAINING

Since the events at Fukushima, there has been an increased interest to expand current simulator capability to address severe accidents. The Modular Accident Analysis Program (MAAP), an Electric Power Research Institute (EPRI) owned and licensed computer software, was developed to simulate and study severe accidents. MAAP is an integral code simulating both containment and primary system during severe accidents. FAI has been under contract to improve MAAP models related to BWR primary system, lower plenum, instrument tubes, molten core concrete interaction and others in order to better follow the severe accident at Fukushima.

Simulators can be expanded to cover severe accidents by implementing the MAAP code into the existing simulator. This implementation using MAAP4 was done for Krsko in Slovenia and Ulchin in Korea. MAAP5 was implemented for Daya Bay in China and Kori in Korea.

MAAP5 PWR code can be a good thermal hydraulic engine for PC based simulator for severe accident training.

The MAAP5 PWR code is the latest generation of MAAP and it implements new models to calculate forced and natural circulation inside a reactor coolant system (RCS) with more detailed nodalization, point kinetic and 1-D neutronics models, features to address details of new advanced reactor designs such as AP1000 and EPR, and improved containment models.

Improvements were also made to include a steam header model with detailed steam dump logic so that the code can calculate initial RCS and steam generator responses after a reactor scram. In addition, MAAP5 has improved models for shutdown states such as modeling nozzle dams in the RCS loops, mid-loop operation, and reactor head open with the vessel submerged under the refueling water pool.

MAAP5 code can also calculate the ANS-3-5 transients required for simulators. These transients include a manual reactor trip, simultaneous trips of feed water pumps, simultaneous closure of all MSIVs, trip of any single reactor coolant pump (RCP), loss of coolant accidents, main steam line break, maximum power ramp, and maximum design load rejection.

MAAP5 FEATURES

The MAAP5 code enhancements build on the substantial MAAP4 code enhancements to assess the approach to, and progression of, severe MAAP Services and PWR RCS Coolant Loop Viewaccidents in BWRs and PWRs operated in the western hemisphere and extends these to the plant specific features that are presently represented in the state-of-the-art models for Probabilistic Risk/Safety Assessments (PRAs/PSAs), Severe Accident Management Guidelines (SAMGs), Emergency Operating Procedures (EOPs) and full scope control room simulators. 

The substantial enhancements in the current MAAP5 code are summarized below: 

  • Natural circulation flows in the core, the hot leg and steam generators, before and after the core is uncovered
  • Improved computation to handle a greater range of transients with one-dimensional and point kinetics neutronics models
  • Increased capability for Steam Generator Tube Ruptures (SGTRs), Main Steam Line Breaks (MSLBs), Loss-Of-Coolant Accidents (LOCAs) analyses which require better steam generator models
  • Improved models for the lower plenum debris pool response including detailed metal layer to wall heat transfer and heavy metal layer formation in the bottom of the lower plenum.  It also has a detailed ex-vessel heat transfer model (nucleate boiling and critical heat flux as a function of azimuthal angle) for in-vessel retention evaluations
  • A spent fuel pool model capable of modeling severe accidents in a spent fuel pool.  The model is capable of calculating fuel uncovery, spent fuel heat up and degradation, Zr oxidation, hydrogen combustion events, Zr fires due to Zr + air interactions, etc.
  • The PWR Reactor Coolant System (RCS) models have continued to advance as the requirements for MAAP have evolved.  

The focus of the MAAP5 RCS model is to provide a fast-running, best-estimate representation of plant response to all types of plant accident conditions.  The goal is to consistently describe the physical processes associated with the integral system response to plant upset conditions, especially those that can progress to severe accident conditions.  MAAP5 models are not intended to replace other software codes that deal with large break guillotine ruptures, which require analysis of the rapid flow reversal within the core, however, it is intended to provide a best estimate description of the core, reactor coolant system, steam generators and containment needed for engineering assessments, including best estimate evaluations of operator procedures and success criteria for PSAs.

Of particular note in MAAP5 is the ability of the primary system model to accommodate independent coolant loop response for PWR designs which can have 1, 2, 3 or 4 steam generator loops.  MAAP5 models the responses of each steam generator (including the number of tubes that are plugged in each generator) depending on the feedwater flows to the generators and their individual steaming rates.  This multiple generator representation can assess the mid-loop operation status for steam generator repair scenarios.

MAAP5 contains both a point-kinetics and a one-dimensional core neutronics model.  As part of this, MAAP5 models the boron distribution within the RCS.  This capability is combined with natural circulation flows due to density differences between 1) the core and the downcomer, and 2) the hot leg and cold leg side of the steam generator tubes.  This provides an integral representation of the RCS and core response when natural circulation flows are important, such as for Anticipated Transient Without Scram (ATWS) conditions.

The MAAP5 containment model builds on the MAAP4 Generalized Containment Model.  It extends the capabilities into several new applications, including Design Basis Accident (DBA).  This model has been extensively benchmarked with containment experiments with the full scale tests from HDR (PWR) and Marviken (BWR) being some of the most important.  Not only do these tests represent the containment response to DBA conditions, but the HDR tests also demonstrate the conditions that would 1) cause hydrogen stratification and 2) the conditions that would result in global mixing to eliminate stratification. 

The MAAP5 containment model has been enhanced so that it can be used for Design Basis Accident (DBA) analyses.  This model will support either a single node or multi-node analysis.  It can also be used to assess the margin in the DBA calculation that results from assuming no contribution from forced convection as a result of the pipe break.  The model also addresses the thermal resistance due to paint layers on the walls.  The containment model includes all the fission product isotopes needed to perform Alternate Source Term (AST) evaluations which are coupled with MAAP5 such that the in-plant and ex-plant doses and dose rates can be calculated in a single run.  It also contains models for the aerosol transport and deposition mechanisms to assess the retention capabilities.

The containment model also includes hydrogen burn models to assess the extent of localized burning (for those conditions where the containment atmosphere is not inerted) that could occur in the containment for severe accidents.  Thus MAAP5 can be used to evaluate the equipment survivability envelope for such conditions.

Additionally, the containment model now has the capability to model all aspects of an accident where the integrity of the spent fuel pool can be challenged.  The MAAP5 code can calculate the time to boil away the pool water inventory, model the heatup and relocation of the spent fuel, the potential for the release of hydrogen from the spent fuel cladding due to Zr oxidation (due to steam and air) and the potential for any type of hydrogen combustion event in the spent fuel pool room/enclosure. 

 

MAAP5 BENEFITS

The MAAP5 software package provides engineers with a tool to rapidly evaluate the progression of accidents in terms of the reactor core (is the MAAP Services and MAAP5 Primary System Nodalization Schemefuel damaged or not?), the containment (is containment integrity being challenged?) and radiological consequences (do the dose rates inside the plant or in the population areas present concerns in terms of taking precautionary measures such as shelter or evacuation?).

MAAP5 also can model the progression of an accident in a plants’ spent fuel pool.  The MAAP5 code can calculate the time to boil away the pool water inventory, model the heatup and relocation of the spent fuel, the potential for the release of hydrogen from the spent fuel cladding due to Zr oxidation (due to steam and air) and the potential for any type of hydrogen combustion event in the spent fuel pool room/enclosure.  Due to the rapid computation speeds and its capability to model all types of reactor transients, Loss-Of-Coolant Accidents and loss of AC/DC power events (SBOs), MAAP5 is a powerful code that can be used in the development of a plant’s accident management strategies.

MAAP5 EXPERIENCE/SOLUTIONS

Originally developed by Fauske & Associates, LLC (FAI) as part of the Industry Degraded Core Rulemaking (IDCOR) program, FAI has developed and maintained the code under the sponsorship of the Electric Power Research Institute (EPRI) and the MAAP Users Group (MUG).

MAAP5, and its predecessor MAAP4, have been used solely by the nuclear industry throughout the world for more than two decades as an engineering tool for severe accident analysis and associated severe accident phenomena, including hydrogen generation and combustion, direct-containment heating, rapid pressurization due to steaming, core concrete interactions, fission product releases, transport and deposition etc.  The MAAP code is also used extensively in the PRA/PSA arena as well for success criteria evaluations, human reliability analyses (HRA) and Level II source term evaluations etc.

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MAAP4 Hot Leg and Lower Head Failure Benchmarking

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