severe accidents and seismic issues in lead-cooled · pdf fileearthquakes occur when stresses...

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






Concluding remarks


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.



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.



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.



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.


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.


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.



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).



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




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.


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.



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


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.



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.



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).



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).



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 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.






Concluding remarks


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.


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


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.


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?






Concluding remarks


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


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


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


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


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



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


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).







Concluding remarks


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


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


Von Mises stress intensity distribution inside the RV.


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.



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


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


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).



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


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.


R. Lo Frano, G. Forasassi, Nuclear Engineering and Design Vol.246 (2012) pp. 423-433


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% )






Concluding remarks


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


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


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


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


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


Thanks for your attention

severe accidents and seismic issues in lead-cooled · pdf fileearthquakes occur when stresses...

Documents