2011-04-06 – NRC – Severe Accidents and MELCOR Code

April 6th, 2011 - NRC - Severe Accidents and MELCOR Code

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Wednesday, April 06, 2011 9:06 AM
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Severe Accidents and MELCOR Code.pdf
DB 139 of 696
Information Sheet: Severe Accidents & MELCOR Code, Dr. Hossein Esmaili (NRC/RES/DSA)
The Risk
The risk to the public by nuclear power generation arises
if accident progress to the point where fuel degradation
occurs, and large quantities of radioactive materials are
released into the environment. The NRC has invested
heavily in the investigation of severe reactor accidents
and has developed computer codes for the analysis of
severe accident phenomena and progression. It is
essential to the mission of the NRC that it possesses
expertise on severe accident phenomenological
behavior and a quantitative predictive capability for
simulating the response of nuclear power systems to
severe accidents. The role of such expertise and
analytical capability is potentially wide ranging in our
regulatory environment including the transition to a more
risk informed regulatory framework and to the study of
vulnerabilities of nuclear power plants.
MELCOR Severe Accident Code
The MELCOR code is a fully integrated, engineering-level
computer code whose primary purpose is to model the
progression of postulated accidents in light water reactors
as well as non-reactor systems (e.g., spent fuel pool and
dry cask). MELCOR is a modular code consisting of
three general types of packages: (a) basic physical
phenomena (i.e., hydrodynamics – control volume and
flow paths, heat and mass transfer to structures, gas
combustion, aerosol and vapor physics); (b) reactorspecific
phenomena (i.e., decay heat generation, core
degradation, ex-vessel phenomena, sprays and
engineering safety systems); and (c) support functions
(thermodynamics, equations of state, material properties,
data handling utilities, equation solvers). These
packages model the major systems of a nuclear power
plant and their associated interactions. MELCOR 1.8.6
(Fortran 77) was released in September 2005, and the
code modernization effort resulted in the release of
MELCOR 2.0 code (Fortran 95) in September 2006. The
latest version (MELCOR 2.1) was released to the NRC in
July 2007.
The Needs
Severe accident competency will be needed to evaluate
new generic severe accident issues and to address risk
informed regulatory initiatives and operating reactor
issues associated with plant changes, as in the case of
steam generator tube integrity.
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Licensees will continue to pursue plant modifications that
require assessment of incremental risk impacts that will
necessitate analysis of severe accident related
phenomena. Licensees in many cases rely on industry
codes, such as MAAP to analyze plant behavior and risk
The NRC will need capability for the foreseeable future to
independently evaluate behavior about which there is still
considerable uncertainty around the best estimate and to
assess phenomena which are regarded quite differently
by various analysts.
MELCOR represents the current state-of-the-art in severe
accident analysis, which has accumulated through NRC
and international research performed since the accident at
Three Mile Island in 1979. The code is fully integrated for
the analysis of severe accident phenomena and
progression. These predictions are required as part of
licensing issues in new reactors and operating reactors’
modifications process. Future needs will include
development and implementation of new and improved
models to predict the severe accident behavior of
advanced non-light water reactor designs.
DB 140 of 696
Information Sheet: Severe Accidents & MELCOR Code-Page 2, H. Esmaili (NRC/RES/DSA)
The improved understanding of phenomenological
behavior and modeling in severe accidents and their
implementation in MELCOR has had a direct impact on
the analytical methods and criteria adopted for design
basis accidents (e.g., source term research and the
revised source term). It is anticipated that the
development of best estimate severe accident models in
the future will improve the licensing evaluation models.
The development of best estimate models reveals,
quantitatively, margins in existing models.
Activities associated with the development, assessment,
and application of MELCOR include:
3 Safety analysis and risk decision-making
o Revision of NRC’s alternative source term
(NUREG-1465) for high-burnup fuel and
mixed oxide (MOX) fuel.
o New reactor certification (AP-1000, ESBWR,
Experimental analyses and code validation activities
> NPP beyond-design-basis accidents
> Aerosol transport and deposition in steam generators
during bypass accidents
Risk of severe accident induced steam generator
tube rupture
Effects of air ingress on fission product release
Vulnerabilities of spent fuel pool to accidents
State-of-the-art consequence analysis
National laboratories, universities (e.g., Texas A&M), and
international organizations (e.g., Paul Scherrer Institute-
Switzerland, Institut de Radioprotection et de SOret6
Nucl6aire (IRSN) – France) are involved in the MELCOR
code development effort.
Examples of international collaborations that
resulted in MELCOR improvement include:
(1) USNRC Cooperative Severe Accident Research
Program (CSARP)
(2) MELCOR Code Assessment Program (MCAP)
(3) Institut de Radioprotection et de SOrete Nucl6aire of
France: Ph6bus-FP, VERCORS, and follow-on
program (Ph~bus-STSET) – fission product releases
and degradation of U0 2 fuel (including burnup
>40 GWD/Mt) and MOX fuel under severe accident
conditions, and the effects of air ingress on core
degradation and fission product release. Results are
used to validate the NUREG-1465 source term and
MELCOR code.
(4) German QUENCH
investigating overheated fu(
experiment Program,
(5) ARTIST – Paul Scherrer Institute (Switzerland): To
investigate experimentally the potential mitigation of
radioactive material releases through secondary side
of a steam generator. Results from this research
would allow the NRC to decide whether improved
source term bypass models are needed.
(6) OECD MCCI program – Argonne National Laboratory
(USA): Separate effects experiments to further
address the ex-vessel debris coolability issue. The
results will be used to develop coolability models for
incorporation into severe accident codes.
(7) CSNI Behavior of Iodine Behavior (BIP) – Nuclear
Energy Agency: Committee on the Safety of Nuclear
Installations (France): Experimental investigations of
behavior of iodine in containment during post severe
accident conditions for computer code model(s)
development and validation. BIP will provide
experimental data to model iodine behavior in
containment. It addresses uncertainties related to
iodine behavior (especially with respect to iodine
interactions with paints in containment). With
complementing testing at Atomic Energy of Canada
Limited (AECL) and at IRSN, the state-of-the-art on
modeling of iodine behavior in the containment can
be advanced and quantified. Adequate modeling of
iodine behavior in the containment is crucial in
determining the need for pH control in containment
sump. The proposed research will complement the
on-going IRSN of France Ph6bus-FP and follow-on
program Ph6bus – Source Term Separate Effects
Test Project.
For More Information
Contact Hossein Esmaili at 301-415-6084
NOTE: Availability of international experimental data
is determined by individual international program
and not by NRC. \3,U.S.NRC
DB 141 of 696