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KIT – University of the State of Baden-Württemberg and
National Large-scale Research Center of the Helmholtz Association
Institute for Nuclear and Energy Technologies
www.kit.edu
Severe accidents and seismic issues in lead-cooled systems
D.Pellini, W.Maschek
SILER Kick-off Meeting, V erona, May 22-23th, 2011
Institute for Nuclear and Energy Technologies 2
Outline
General considerations on seismic issues
Design Extended Conditions & Severe Accidents
Studies performed in the past
Lead cooled reactors & seismic issues
Concluding remarks
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General considerations on seismic issues
Design Extended Conditions & Severe Accidents
Studies performed in the past
Lead cooled reactors & seismic issues
Concluding remarks
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General considerations on seismic issues Seismic Siting Criteria
Earthquakes occur when stresses in the earth exceed the strength of a rock mass,
creating a fault or mobilizing an existing fault.
3 kinds of fault‘s movements:
1. lateral (a strike/slip fault)
2. vertical (a thrust or reverse fault)
3. combination of the two movements
The fault’s sudden release sends seismic shock waves through the earth that have two
primary characteristics:
1. Amplitude - a measure of the peak wave height
2. Period - the time interval between the arrival of successive peaks or valleys.
The seismic wave’s arrival causes ground motion.
The intensity of ground motion depends primarily on three factors:
1. Distance from the source (also known as focus or epicenter)
2. Amount of energy released (magnitude of the earthquake)
3. Type of soil or rock at the site.
In general, for a given magnitude earthquake, the shallower the focus, the stronger the
wave will be when reaching the surface. In addition, the intensity of ground shaking
diminishes with increasing distance from the earthquake focus.
Sites with deep, soft soils or loosely compacted fill will experience stronger ground
motion than sites with stiff soils or rock.
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General considerations on seismic issues
Earthquake Magnitude
Earthquake Magnitude is a measure of the strength of the earthquake as determined
from seismographic observations and records (e.g., Richter Local Magnitude, Surface
Wave Magnitude, Body Wave Magnitude, and Moment Magnitude).
Currently, the most commonly used magnitude measurement is Moment Magnitude (M),
which accounts for
1. strength of the rock that ruptured
2. area of the fault that ruptured
3. average amount of slip.
Moment is a physical quantity proportional to the slip on the fault times the area of the
fault surface that slips. It relates to the total energy released in the earthquake and can
be estimated from seismograms and from geodetic measurements.
Moment Magnitude provides an estimation of earthquake size that is valid over the
complete range of magnitudes, a characteristic that was lacking in other magnitude
scales, such as the Richter scale.
The common measure of an earthquake’s magnitude (M) refers to the logarithmic
Richter scale, thus an M 7.0 earthquake has an amplitude that is ten times larger than
an M 6.0, but releases 31.5 times more energy than an M 6.0 earthquake.
However, the term Richter Scale is so common in use that scientists generally just
answer questions about “Richter” magnitude by substituting moment magnitude
without correcting the misunderstanding.
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General considerations on seismic issues
Intensity
The intensity of an earthquake is a qualitative assessment of effects of the earthquake at
a particular location.
The assigned intensity factors include observed effects on humans, on human built
structures, and on the earth’s surface at a particular location such as in the Modified
Mercalli Intensity (MMI) scale, which has values ranging from I to XII in the order of
severity.
Greater magnitude earthquakes are generally associated with greater lengths of fault
ruptures.
The length of the fault break, however, is not directly proportional to the energy
released.
The induced amplitude of acceleration (g) does increase with increasing magnitude (M).
Various methods developed relate the magnitude of an earthquake to the
amplitude of acceleration it induces, and different methods may result in
significant variations in results.
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General considerations on seismic issues
In the late 1940s, structural engineers in USA began considering the seismic-based
shear forces that structures must resist.
To supplement their design calculations, they referred to the Seismic Zone Map
published by the Uniform Building Code (UBC, in Figure B-1).
The UBC map divided countries into several distinct seismic zones representing
various degrees of seismic risk.
The map expressed peak ground acceleration as the decimal ratio of the acceleration
due to gravity (g) that applied to a Maximum Credible Earthquake (MCE) and an
Operating Basis Earthquake (OBE).
UBC defined a maximum credible earthquake as producing the greatest level of
ground motion at a certain site.
An operating basis earthquake was defined as the greatest level of ground
motion likely to occur during the economic life of a structure.
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General considerations on seismic issues Safe Shutdown Earthquake Condition
In 1973, the concept of the “safe shutdown earthquake” (SSE) was introduced in Title 10 Part
100 of the Code of Federal Regulations (10 CFR 100), Appendix A—Seismic and Geologic
Siting Criteria for Nuclear Power Plants.
The NRC defines the Safe Shutdown Earthquake as the maximum earthquake in which certain
structures, systems, and components, important to safety, must remain functional.
Under an “operating basis earthquake,” the reactor could continue operation without undue
risk to the safety of the public.
Ground motion at any specific location, such as a nuclear plant site, depends on the
earthquake source, magnitude, distance to the source, and the attenuation (dampening)
caused by rock and soil characteristics.
A nuclear power plant responds to an earthquake depending on how its individual structures,
systems, and components resonate, or vibrate, with the ground shaking.
Heavier and more massive structures resonate at lower frequencies, while light components
resonate at higher frequencies.
During an earthquake, ground motion transmits vibrations to a nuclear power plant’s
foundation and structure. Vibrations cause back-and-forth acceleration of structures, systems
or components that is measured relative to the earth’s gravitational acceleration constant (g).
Both vertical and horizontal components of ground acceleration place loads, or stresses, on a
nuclear power plant’s structure.
Peak Ground Acceleration (PGA) is a measure that has been widely used in developing nuclear
power plant “fragility estimates,” which represent the sensitivity of nuclear plant structures,
systems, and components (SSCs) to the inertial effects of acceleration during ground shaking.
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Cumulative Absolute Velocity
Structural damage to nuclear power plants occurs when the cumulative effects of
ground acceleration (seismically induced vibrations) cross a certain threshold.
The Electric Power Research Institute (EPRI) developed the concept of “cumulative
absolute velocity” (CAV) in 1988 as an index for indicating the onset of structural
damage from the cumulative effects of ground acceleration.
The threshold between damaging and non-damaging earthquakes (for well-designed
buildings) conservatively occurs at ground motions with cumulative absolute velocities
(CAV) greater than 0.16 g-seconds.
In simple terms, CAV is the sum of various ground acceleration frequencies (measured
in terms of g) and the duration of their acceleration (measured in seconds).
An example of this phenomenon is a wire coat-hanger that breaks from metal fatigue
after being rapidly bent multiple times.
Experimental and empirical seismic data have provided insights into the behavior of
different structures under various acceleration and shaking conditions.
For example, welded steel piping at nuclear power plants rarely failed when peak
ground accelerations remained below 0.5g.
Other types of structures exhibit different behaviors.
Engineers design the various plant structures to withstand a certain severity of
earthquake and estimates of ground shaking specific to each plant site.
General considerations on seismic issues
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The maximum vibratory accelerations of the Safe Shutdown Earthquake must take into
account the characteristics of the underlying soil material in transmitting the earthquake
induced motions at the various locations of the plant’s foundation.
Various plant structures, depending upon their elevation above the foundation,
vibrate at different frequencies during an earthquake.
Vibrations in the range of 1 to 10 Hz are particularly problematic, because a wide range
of structures are susceptible to damaging resonance at those frequencies.
These accelerations and the corresponding shaking frequencies are factors in the
Probabilistic Seismic Hazard Analysis (PSHA, discussed below).
The full seismic spectrum often can be characterized by two intervals:
1. peak ground acceleration (PGA)
2. spectral acceleration (SA) averaged between 5 and 10 hertz (Hz).
General considerations on seismic issues
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Seismic Hazard
Seismic hazard is defined as the probable level of ground shaking associated with the
recurrence of earthquakes.
The assessment of seismic hazard is the first step in the evaluation of seismic risk
obtained by combining the seismic hazard with local soil conditions and with
vulnerability factors (type, value and age of buildings and infrastructures, population
density, land use).
Frequent, large earthquakes in remote areas result in high seismic hazard but pose no
risk but on the other , moderate earthquakes in densely populated areas entail small
hazard but high risk.
Minimization of the loss of life, property damage, and social and economic disruption
due to earthquakes depends on reliable estimates of seismic hazard.
National, state and local governments, decision makers, engineers, planners,
emergency response organizations, builders, universities, and the general public
require seismic hazard estimates for land use planning, improved building design and
construction (including adoption of building codes), emergency response
preparedness plans, economic forecasts, housing and employment decisions, and
many more types of risk mitigation.
Two approaches are followed:
1. Deterministic Seismic hazard
2. Probabilistic Seismic hazard
General considerations on seismic issues
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General considerations on seismic issues
Deterministic Seismic Hazard Analysis
During the 1960s and 1970s designs for nuclear power plants granted construction
begin to apply a deterministic approach to seismic design based on site-specific
investigations of local and regional seismology, geology, and geotechnical soil
conditions to determine the maximum credible earthquake from a single source (fault).
Deterministic Seismic Hazard Analysis (DSHA) attempted to quantify the effects of a
maximum credible earthquake based on known seismic sources sufficiently near the
site and available historical seismic and geological data to estimate ground motion at
the plant site.
Appendix A to 10 CFR 100 requires an investigation of fault and earthquake
occurrences to provide the basis for determining a safe shutdown earthquake.
Appendix A to 10 CFR 100 notes the limitations for basing seismic design criteria on
literature reviews of geophysical and geologic information, and requires
supplementing the investigation with studies for vibratory ground motion, evidence of
surface faulting, and evidence of seismically induced floods and water waves that
have or could have affected the site.
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General considerations on seismic issues Probabilistic Seismic hazard
The basic elements of modern probabilistic seismic hazard assessment can be
grouped into four main categories:
1. Earthquake Catalogue: the compilation of a uniform database and catalogue of
seismicity for the historical (pre-1900), early-instrumental (1900-1964) and
instrumental periods (1964-today).
2. Earthquake Source Model: the creation of a master seismic source model to
describe the spatial-temporal distribution of earthquakes, integrating the
earthquake history with evidence from seismotectonics, paleoseismology, mapping
of active faults, geodesy and geodynamic modeling.
3. Strong Seismic Ground Motion: the evaluation of ground shaking as a function of
earthquake size and distance, taking into account propagation effects in different
tectonic and structural environments.
4. Seismic Hazard: the computation of the probability of occurrence of ground
shaking in a given time period, to produce maps of seismic hazard and related
uncertainties at appropriate scales. It depicts the levels of chosen ground motions
that likely will, or will not, be exceeded in specified exposure times.
Hazard maps commonly specify a 10% chance of exceedance (90% chance of non
exceedance) of some ground motion parameter for an exposure time of 50 years,
corresponding to a return period of 475 years.
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General considerations on seismic issues
DSHA had based peak ground acceleration (PGA) on a single earthquake source, PSHA uses up-to-date interpretations of earthquake sources, earthquake recurrence, and strong ground motion estimates to estimate the probability of exceeding various levels of earthquake-caused ground motion at a given location in a given future time period.
It quantifies a site’s seismic hazard characteristics from seismic hazard curves or “response spectra” developed in part by identifying and characterizing each seismic source in terms of maximum magnitude, magnitude recurrence relationship, and source geometry.
Under 10 CFR 100.23 (Geologic and Seismic Siting Criteria), designs for new nuclear power plants will base their Safe Shutdown Earthquake on Probabilistic Seismic Hazard Analysis (PSHA).
Probabilistic Seismic hazard (cont’d)
The methodology has also found widespread use in U.S. engineering practice for nonnuclear structures. PGA is the most commonly mapped ground motion parameter because current building codes that include seismic provisions specify the horizontal force a building should be able to withstand during an earthquake
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Site-specific ground motion response spectrum
Each earthquake produces a unique sequence of ground motions
(accelerations) that may last several seconds or longer.
The record of ground motion, captured on an accelerograph, appears as a
jagged-shaped line that represents the peak values of acceleration/de
acceleration.
The ground motion response spectrum represents the range of multiple
earthquake records.
General considerations on seismic issues
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Seismic Hazard Assessment in the European-Mediterranean region
Over the past years three main project frameworks have aimed at improving regional
seismic hazard assessment in the European-Mediterranean region, by integrating
earthquake catalogues, seismic source zoning and hazard assessment.
The Global Seismic Hazard Assessment Program (GSHAP) produced the first seismic
hazard map for the European-Mediterranean region as part of the Global Seismic Hazard
Map based on the compilation and assemblage of hazard results obtained independently
in different test areas and multinational programs (Adria, Ibero-Maghreb, Central-
Northern Europe, Fennoscandia, Turkey and Greece, Caucasus, Near East, the Balkans).
The International Geological Correlation Program project n.382 Seismotectonics and
seismic hazard assessment of the Mediterranean basin (SESAME) developed in year
2000 the first integrated seismic source model and homogeneous hazard mapping for
the Mediterranean region.
General considerations on seismic issues
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Seismic Hazard Assessment in the European-Mediterranean region cont’d)
Finally, the European Seismological Commission (Working Group on Seismic Hazard
Assessment) has completed the first unified seismic source model and seismic hazard
mapping for Europe and the Mediterranean.
The unified seismogenic source model for the whole Mediterranean region consists of a
total of 463 seismic sources (455 shallow and 8 intermediate-depth). Each source is
characterized by seismicity parameters in terms of earthquake activity rates and
maximum magnitude.
The ESC-SESAME model for the European-Mediterranean region allows the generation
of hazard maps expressing ground motion in different parameters, for different soil
conditions and probability levels through a homogeneous computational procedure.
This map is computed using the PGA attenuation laws of Ambraseys et al. (1996),
Musson (1999), and Papaioannou and Papazachos (2000) and the areas not covered by
the ESC-SESAME seismic source model (Iceland and Russia) are taken from the Global
Seismic Hazard map (GSHAP).
General considerations on seismic issues
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European seismic hazard map
Horizontal peak ground acceleration seismic hazard map representing stiff site conditions for an
exceedance or occurrence rate of 10% within 50 years for the mediterranean region.
Map colors are chosen to represent roughly the actual level of hazard. In particular, white to green
correspond to low hazard (0-8% g) yellow and orange to moderate hazard (8-24% g) reds to brown
high hazard (> 24% g).
General considerations on seismic issues
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These “safety-related” structures, systems, and components are those
necessary to assure:
• capability to maintain the reactor coolant pressure
• capability to shut down the reactor and maintain it in a safe
condition
• capability to prevent or mitigate the consequences of
accidents which could result in potential offsite radiation
exposures.
General considerations on seismic issues
General design criteria for nuclear power plants require that structures and
components important to safety withstand the effects of earthquakes,
tornados, hurricanes, floods, tsunamis, and seiche waves without losing the
capability to perform their safety function.
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General considerations on seismic issues
Design Extended Conditions & Severe Accidents
Studies performed in the past
Lead cooled reactors & seismic issues
Concluding remarks
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Design Extended Conditions
Design Extended Conditions (DEC) are defined as a specific set of accident sequences
that must be selected on deterministic and probabilistic basis going beyond Design
Basis Conditions (DBC).
DEC include:
1. Complex Sequences
Correspondence to sequences considering failures of mitigating systems beyond Design
Basis Condition (DBC) together with the failure of one or several mitigating systems.
Number of failed systems defined according to safety objectives (probabilistic methods or
preliminary design methods).
Design additional systems and/or to adapt existing systems in order to satisfy safety
objectives (Severe core damage prevention).
2. Limiting Events
Accidents conditions representing cases of particular fault types important for license
purposes.
Postulation based on technology specific risks (e.g. local fault in the core, common fault
failure).
Studies aimed at showing no occurrence of cliff edges (e.g. no significant core degradation).
3. Severe accidents
Accidents corresponding to situations with a significant core degradation.
Target: no need of protective measures for people living in the proximity of the plant.
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Expansion phase In case of severe accident : discharge of
molten material from core and acceleration of surrounding coolant; redistribution of granulated fuel
Important for work energy potential and mechanical structure load assessment after severe accident
Upper core and vessel structures & behavior to be known (impact on mitigation)
Several experimental campaign have been set up in order to study in depht the phenomenology of the expansion phase
Expansion phase phenomenology
Possible Damage: energetic sodium pool impact (left)
mild sodium pool impact (right)
Severe accident consequences
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Potential Initiators of Severe Accidents in HLM reactors
“Classical transients” for fast reactors (whole core involvement):
• ULOF
• UTOP
• ULOHS
Are they enough for HLM reactors?
Sequences that needs particular attention for HLM
TIB (Total Instantaneous Blockage) • Bounding case treated in SFR
• Pin disruption
• Propagation
• Detection & scram
• Prevention
SGTR (Steam Generator Tube Rupture)
• CCI (Coolant-coolant Interaction)
• Sloshing
• Potential gas entry in core region
The analysis of initiators should fully cover accidents’ scenarios
and phenomena involved.
Simulation tools have a crucial role in studying accidents’ sequences.
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Safety issues for Gen IV Reactors
Safety goals:
Prevention: reduction of likelihood of accident occurrence
• Assessment of “black swans”
• Assessment of scenario “to be practically excluded”
• Assessment of effectiveness of the safety countermeasures against accidents
• Assessment of adequacy of simulation tools
Mitigation:
• Complex sequences & Limiting Events: no need of protective measures for people living in the
proximity of the plant.
• Whole core damage: no permanent relocation, no need for emergency evacuation outside the proximity of the plant, limited sheltering, no long term restriction in consumption of food.
Design measures for Generation IV reactor concepts that should be taken under Design
Extended Conditions (DECs) must include countermeasures for external events, too.
Core cooling under long term loss of electric power or failure of auxiliary systems shall
be ensured by using diverse decay heat removal systems.
Alternative heat removal measures under core damage situations have also to be
considered.
The strong influence of human factor and a high grade of complexity affects scenarios
and accidents’ development R&D in severe accident play a crucial role in all kinds
of reactors.
Gen-IV systems’ safety requirement must go deeper than the current safety
approach for LWR?
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General considerations on seismic issues
Design Extended Conditions & Severe Accidents
Studies performed in the past
Lead cooled reactors & seismic issues
Concluding remarks
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Fast reactors issues Upper Core Structure
CR Scram delay
Core Barrel SA Vibrations
CR Scram delay
Core Support Plate Bounding of SA
CR Drawing out
Quick Response
(a few seconds)
Reactor Trip
CR Scram
Pump trip
• Dominant reactivity
component of core damage
accident
• Probability distribution of
energy release
• Evaluation of the accident
effect
Under exceeding condition of
Design Basis
Earthquake
Ground Motion
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Subassemblies Vibration
Particular of
Control Rod
Horizontal core cross section
CR inner&outer core Blanket
Dummy
Monju’s Layout
&
Core Cross Sections
Vertical cross section
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Subassemblies Vibration
• Subassemblies stand on core
support plate by own weight
• Small gap between neighboring
subassemblies
• Vibration with impact on
wrapper tube pad
• Positive reactivity induced by
reduction of the distance
between neighboring
subassemblies
Subassemblies
vibration
Subassembly Flow Path
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PEC Analysis PEC reactor scheme
Core cross
section
Contraction of PEC central row
core elements calculated at SSE
for the initial underformed
geometry (left) and the end of life
for deformed geometry at ½ SSE
(right).
Reactivity insertion C. Artioli, F. Cecchini, P. Corticelli, R. Di Franceses, M. Forni, A. Martelli,
P. Montanelli, J. Me Loughlin, P.G. Muratori,“Evaluation of the neutronic-
seismic interaction effects in the PEC Fast Reactor Core Analysis” Proc.
Int.Topic Meeting on Fast Reactor safety, Knoxville,April 21-25,1985
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Sloshing Analyses in Sodium Cooled Fast Reactors
Extensive theoretical (codes) and experimental work performed for past SFR
projects
HLM : higher density and lower compressibility
Y.W. CHANG, Nuclear Engineering and Design 106 (1988) 19-33
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Sloshing Analyses in Sodium Cooled Fast Reactors
Examples of experimental work :
A. SAKURAI, Nuclear Engineering and Design
Vol.113 (1989) pp. 423-433 R. Aziz Uras, ANL/RE/CP-85929
Shaking Table Tests
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Sloshing – Experimental campaign
in KIT (1992)
Maschek, W., Roth, A., Kirstahler, M., Meyer, L., “Simulation Experiments
for Centralized Liquid Sloshing Motions,” KfK report, KfK5090 (Dec. 1992).
Sloshing Analyses in Sodium Cooled Fast Reactors
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General considerations on seismic issues
Design Extended Conditions & Severe Accidents
Studies performed in the past
Lead cooled reactors & seismic issues
Concluding remarks
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Seismic issues in Lead cooled reactors
Stress intensity distribution resulting
from Lead motion coupled to the seismic
wave propagation might damage structures
thus jeopardizing their resistance to dynamic
loads on the reactor vessel and internal
components.
Seismic events in Lead cooled reactors
may lead to:
• Sloshing
• Buckling (from Sloshing)
• Thermal loads due to the residual
decay heat
Seismic analysis is aimed at:
• assessing seismic waves propagations
inside the containment building
• assessing the influence of isolators
• assessing structural effects due to dynamic
loads on the reactor vessel and on the main
internals
ELSY Reactor
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Sloshing in Lead cooled reactors
SIMMER III Model for EFIT
Conditions:
• Pressure in SG tube: 147 bar
• Pressure in HLM: ~ 6 bar
• Sub-cooled water: 335 °C
• Rupture type: Guillotine crack Late phase of a SGRT accident with steam entry into the
core
Example: SGTR - EFIT
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Von Mises stress intensity distribution inside the RV.
Sloshing in Lead cooled reactors
R. Lo Frano, G. Forasassi, Energy Vol.36 (2011) pp. 2278-2284
Input horizontal acceleration for isolated CB structure.
Fluid motion at the beginning of sloshing (a) and after some seconds (b).
Example: ELSY- EFIT
Inner structures
influence the
fluid waves
motion acting as
baffles and thus
reducing lead
mass waves
size.
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Sloshing in Lead cooled reactors
Buckling Problem
Buckling phenomenon occurs when most of the strain energy,
which is stored as membrane energy, can be converted into
bending energy required by large deflections.
Phenomenon can be connected to sloshing.
Seismic loading, due to LBE sloshing effect, may produce
stresses exceeding the allowable limits in localized parts of the
reactor internals.
• Simulation performed for PDS-XADS
• ~ 2000 tons of LBE
PDS-XADS Scheme
PDS-XADS FEM Model
Buckling Pressure
evaluation
• Internals: 20.37 MPa
• Vessel: 34.4 MPa
Values higher than
sloshing pressure
R. Lo Frano, G. Forasassi, Journal of Achievement in Materials and Manifacturing Engineering, Vol.29, August 2008,pp. 163-166
Internals & RV deformation
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Isolators
Isolation systems are aimed to increase the
fundamental period of structural vibration beyond
the energy containing period of earthquake ground
motions, and to reduce the acceleration transferred
to the structures above.
The energy dissipation level of NPP buildings and
structures can be considerably increased by
suitable isolators systems.
They might decrease the overall structure dynamic
response in terms of accelerations.
In general isolation devices are placed
underground, in order to combine the acting
components horizontally and vertically.
Efficient isolation systems reduce the effects on
structure due to earthquake intensity.
R. Lo Frano, G. Forasassi, Nuclear Engineering and Design Vol.246 (2012) pp. 423-433
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During SSE earthquake event, the reactor should be shutdown and the fuel should remain
coolable.
Evaluation of the heat source term for the high burn up levels is assumed (conservative
condition) aimed at discovering possible problems in structures due to thermal loads.
Thermal Loads due to the residual decay heat
ELSY core assumptions:
• Active zones (n. 11, 12, 13) represent the three
fuel composition having respectively 14.6%,
15.5%, 18.5% Pu reactor grade enrichment; each
zone has an height equal to the active core height
and the volumes involved have values between 5
and 10 m3.
• Lower and upper fuel pins structural zones (n. 9,
7, 5 and 10, 8, 6) include the bottom-plug,
insulator, top-plug, spring, etc.)
• Dummy assembly zone (n. 4) with magnesium
oxide (MgO) used as reflector. Scheme of ELSY core zones for
evaluation of heat source term
Several burn up conditions has been considered for each fuel zone for assessing the
decay power. It has been also assumed to couple the SSE loading to the thermal ones
induced by the decay power density after 1825 effective full power days (efpd) of fuel
irradiation. This values was chosen because nuclear fuel irradiated for 1825 efpd
reaches a burnup of 74.56 GWd/tonHM, that is not so far from the burnup target of a
typical LFR reactor (~100 GWd/tonHM).
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Thermal Loads due to the residual decay heat
Different decay power density behaviours, calculated
immediately after shutdown (central core zone). Transient decay power for the three core active zones
considered.
ELSY model with core region
characterization
R. Lo Frano, G. Forasassi, Nuclear Engineering and
Design Vol.246 (2012) pp. 423-433
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Temperature distribution analysis has confirmed the need of the DHR system which
removes decay heat ensuring thermo-structural integrity of the reactor structures
despite the direct contact of the hot lead coolant with the reactor vessel itself.
A failure of the DHR system would lead to RV temperature increase thus transferring
heat through ELSY reactor in all directions.
Temperature distribution (K), inside RV with (a) and without (b) the action of decay heat removal systems.
Thermal Loads due to the residual decay heat
R. Lo Frano, G. Forasassi, Nuclear Engineering and Design Vol.246 (2012) pp. 423-433
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Decay Heat in LFR with Minor Actinide Load
The failure of decay heat removal was the key issue in the Fukushima. To manage the decay heat problem the natural convection paths have to be guaranteed and also the decay heat levels have to be well known
Of special importance if LFRs should manage waste and burn Minor Actinides (MAs)
Results of a benchmark exercise. Fuel options considered : MOX fuel and ADS-EFIT-type fuel with high, 50% fraction of MAs
The decay heat includes mainly two components: FPs and actinides decay heat
The actinides decay heat is sensitive to the irradiation time and isotopic content of the fuel mainly due to the production of Cm242
For long cooling times the decay heat in fuels with MAs may be several times higher than in MA-free fuels
Decay heat determination after an excursion history & influence on temperature level of the pool and potential for vaporization of fission products and MAs and their redistribution.
0
1
2
3
4
5
6
7
0.1 10 1000 100000 1e+007
Decay h
eat, %
of
fiss
ion p
ow
er
Time, s
Mox 30 daysPu/MA: 50/50 30 days
Mox 500 daysPu/MA: 50/50 500 days
Decay heat for the MOX and Pu/MA: 50/50 fuels
after 30 and 500 days of irradiation (decay heat
values obtained for all fuel cases in this study are
relative to the fission power at operating
conditions, this power being lower than the total
reactor power by a value of the order of 10% )
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General considerations on seismic issues
Design Extended Conditions & Severe Accidents
Studies performed in the past
Lead cooled reactors & seismic issues
Concluding remarks
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World Wide Stress Tests for NPPs Could give
Guidance for Safety Design
Hazards caused by nature
Earthquake
Flooding
Extreme weather conditions
Hazards caused by civilization
air plane crash
terroristic threats
explosions due to gas release
attacks on software systems
incidents in neighboring reactor block
Additional defined beyond design basis postulates
Station blackout
Long term external power supply failure
Loss of ultimate heat sink
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World Wide Stress Tests for NPPs Could give
Guidance for Safety Design (cont’d)
Robustness of preventive systems
Quality of separation of redundant systems
Internal hazards affecting more than one redundant system
Natural hazards to ultimate heat sink
Complicate conditions in emergency cases
destroyed infrastructure incl. communication
emergency management under natural hazards
fission product release
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Potential Lessons that Could be Learned from
Fukushima for Future Designs
Events were not expected neither in their strength nor in their
consequences, leading to a common mode failure and a simultaneous
meltdown of several reactors.
Attention to cliff-edge effects
Special protection to external events
The Fukushima accident evolved in a complex nature
Strengthen analyses of severe accidents to provide adequate preventive
and mitigative measures
Preparation of adequate accident management measures under severe
conditions
Installation of severe accident instrumentation
Important lesson : put a focus on rare initiators, accident routes and
consequences that are neither expected nor have been observed, events
that are categorized under ‘black swans’
New Extended Safety Strategies (e.g. France) : Hardened Safety Core - to
prevent a severe accident, limit its progression, limit massive radiological
release in an accident scenario which would not be mastered and allow
easier the operator to manage emergency situations
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Proposals for Additional Measures (IAEA)
IAEA member states propose an additional layer of protection for
nuclear plants to prevent a severe accident, regardless of the
initiating event:
• Improvement of emergency response and management capabilities
• Improvement of hydrogen explosion control
• Implementation of more robust instrumentation in the reactors and
spent fuel pools
• Implementation of stronger accident mitigation measures
• Prevention of an accident's progression to a situation that results in
fuel damage and melting
• Additional fixed and mobile equipment should be considered to
provide the increased capacity to meet essential functions, such as
delivering power and cooling water
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Final Remarks
Specific for HLM
Identification of routes into severe accident scenarios with core damage
Failure conditions for fuels and clad under HLM conditions
Potential of blockage formation, detection, material deposition, growth
conditions (advanced fuels)
Impact of released fuel & steel after pin failure, e.g. after local blockage
formation, detection, eutectic formation, redistribution in vessel
SGTR phenomena, detection, CCI potential, sloshing phenomena and
impact forces
Impact of internal structures on the evolution of the accident
Code development and validation for phenomena related to severe
accident conditions
Scaling factors
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Thanks for your attention