Transcript Kein Folientitel - Nuclear Energy Agency

OECD/NEA/CSNI-WGRisk
“International Workshop on Level 2 PSA and Severe Accident Management”
Cologne, Germany, March 29-31, 2004
A PROPOSAL TO ASSESS CONDITIONAL PROBABILITY
RANGES FOR PLANT INTERNAL PHENOMENA DURING
CORE MELT SCENARIOS FOR GERMAN LWR
J. Eyink*, T. Froehmel**, H. Loeffler***
*Framatome-ANP GmbH, Erlangen, Germany
**Bundesamt für Strahlenschutz (BfS), Salzgitter, Germany
***Gesellschaft für Reaktorsicherheit mbH (GRS), Koeln, Germany
OECD Workshop on Level 2 PSA and SAM
 Overview on regulatory Guidance for Level 2 PSA in Germany
 Methods for Quantification of Branching Probabilities in APET for German LWR
 Recommendations and Examples
Content
Guide
Basics of the
Periodic Safety Review
Guide
Safety Status
Analysis
Protection goal
oriented structure of
nuclear regulations
- Overview of
fundamental
requirements, 12/96
Guide
Probabilistic
Safety Analysis
Guide
Analysisof
Physical Protection
Methods for probabilistic safety analysis
for nuclear power
plants, 12/96
Data for quantification of
event sequence
diagrams and event
trees, 4/97
Guidance for Level 2 PSA in Germany
Task force
"PSR" of the
Federal Committee for
Nuclear
Energy,
chaired by
BMU
published: 1997
> revision process
draft: 2003
expert groups
chaired by BfS
revision process
draft: 2004
new:
• Level 2 PSA, technical
details on methods and
data
Main objectives of the guidance are:
 to support a systematical assessment of branching probabilities for severe accident
progression event tree analysis for German LWR and
 broadening the information and database for this analysis step of Level 2 PSA
 to specify for which branching probabilities generic, plant-typ specific or plant
specific numbers need to be used
 to reduce the potential of controversial expert views on complex and not well
known severe accident phenomena which might be difficult to resolve in the frame
of Periodic Safety Review-process
APET Branching Probabilities for German LWR
selected phenomena:
 Depressurization of the RCS
 Arrest of core degradation in-vessel
 Molten Core-Water-Interaction (in-vessel steam explosion)
 Hydrogen combustion
 Loss of RPV integrity under high pressure
 Coolability of core debris (ex-vessel)
 Arrest of core-concrete-interaction
 Pressurisation of the containment
description:
 every phenomenon is described in a qualitative way addressing the key physical
and chemical features.
 available methods to treat the problem are presented
 methods are illustrated by examples based on available plant specific PSA
 general recommendations and - if possible - quantitative values are issued on how
to deal with the problem in PSAs
Approach
Phenomenon or
Recommendation
Process
Depressurisation of the Evaluation of the available time
RCS
for the operator action
Evaluation of the success taking
into account previous operator
errors
Rationale
This analysis is necessary
for each PDS.
Also for similar plants a
different mix of scenarios
with different time
progression may contribute
to each PDS
 Calculations of RCS-pressure progression shall be performed with / without
depressurisation for a number of scenarios and results shall be plant-specific
 Determination of the period of time is important to get the points of time for
available information and latest initiation of active primary bleed in order to avoid
RPV failure at high pressure
 Determination of probability of successful operator action depending on time and
various stress-situations by using e.g. SWAIN-data
Depressurisation of the RCS
Phenomenon
or Process
Failure of
components of
the RCS (e.g.
hot leg, SGT) in
case of high
pressure in the
RCS
Recommendation
Rationale
Basis should be plant specific calculations on
surface temperature history. Creep failure can
be either calculated with the thermohydraulic
code, if possible, or general correlation can be
applied considering the proper material.
For SGT at these conditions:
Tube-temperature: 700 – 800 °C
RCS-pressure:
80 bar
following generic values can be applied
5% fractile: 0.01
50% fractile: 0.05
95% fractile: 0.5
The distribution of failure probability considers
pre existing mechanical damages as well as
intact tubes.
The effort devoted
to modelling RCS
failure should be
commensurable to
its importance
 intervalls of temperature and pressure should be used broad enough in order to
compensate considerable unscertainties of models and accident conditions
Depressurisation of the RCS
Phenomenon or Process
Arrest of core degradation
in-vessel
Recommendation
In case of PWR a generic
correlation is given:
if core degradation is less
than 20% and coolability
can stabilised by water
injection RPV integrity can
be assumed.
Rationale
TMI as example;
Similar correlations for
BWR are to be established.
The large degree of
uncertainty does not justify
plant specific investigation
Probability of Core Coolability [%]
 reliability data resulting from level 1
 degree of core degradation is the ratio of
fuel which lost pellet geometry and the
75
entire fuel mass
50
 determination of likelihood of coolability
depending on degree of core degradation
25
by deterministical analyses:
0
- of degradation gradient (dep. on scenario)
- of starting point of water injection
0
20
40
60
80
Degree of core degradation [%]
- of keeping the partially destroyed core
in configuration
Ref.: Löffler, H. et al.: Untersuchung auslegungsüberschreitender Anlagenzustände mittels
100
Ereignisbaumtechnik am Beispiel einer Konvoi-Anlage. BMU-2002-594, November 2002,
ISSN 0724-3316
Arrest of core degradation in-vessel
Phenomenon
or Process
In vessel
steam
explosion
Recommendation
Rationale
For the probability alpha mode failure
of the containment the following
generic values can be applied for the
PWR:
5% fractile: 10-5
50% fractile: 10-4
95% fractile: 10-3
A method is presented that can be
applied to BWR.
The numbers are in line with
international assessments
(e.g. SERG2) and new
experimental findings
(BERDA, ECO)
The low probabilities and the
large uncertainties justify
generic numbers.
Quantification of containment failure probability by comparison of potential
explosion loads and the RPV failure limits and using experimental findings:
 determination of debris mass (incl. thermal energy) which can interact with water in the
RPV-bottom head
 determination of mechanical energy resulting from explosions by using conversion
factors for calculated thermal energy,
 comparison of mechanical energy and RPV failure limits (energy transmission by
accelerated compact water plug)
Molten-Core-Water-Interaction Consequence
Phenomenon
or Process
Failure of the
RPV under
high pressure
(> 80 bar)
Recommendation
Rationale
a) leak size PWR
a) Sandia LHF
2
5% percentile: 0.04 m
experiments as
2
50% percentile: 0.3 m
basis
2
95% percentile: 3.6 m
d) Simplified
b) leak size BWR
methods
2
5% percentile: 50 cm
available in the
2
50% percentile: 0.05 m (10 penetrations)
literature
2
95% percentile: 3.6 m
c) containment rupture by missiles (1 m2)
large leak (>3.6 m2): 90%
small leak (< 3.6 m2): 10%
d) containment rupture due to DCH (1 m2)
plant typ specific calculation required taking into
account Zr oxidation and hydrogen combustion.
Relevant factors of RPV failure mode (PWR/BWR):
 mode of bottom head failure at increased pressure with large or small leak
 wetted or unwetted bottom head
RPV failure and Containment failure modes
Phenomenon Recommendation
or Process
Hydrogen
Plant specific calculations are required in
combustion
order to get:
 The range of hydrogen production (in
total and for each PDS)
 The likelihood of ignition as function of
time taking into account plant specifics
of the electrical installations and the
recombiners
 Plant specific calculations of the
pressure distributions for each PDS in
case of combustion
 Plant type specific analysis of the
structural behaviour of the
containment (failure probability and
mode as function of the pressure)
Hydrogen combustion
Rationale
The large contribution of
containment failure due to
hydrogen combustion to the
risk justifies a partly plant
specific approach;
validated tools for such
calculations are available
Core-concrete-interaction in dry cavity
 It has to be assumed, that core material – without water covering – completely melts
through the basemat and that released gaseous components increase the
containment pressure (PWR, BWR Typ 72). In case of one BWR-type (Typ 69) fire
hazards shall be considered in adjacent buildings (e.g. reactor building). Important
parameters (erosion rate, gas release rate) relevant to containment integrity are to
be estimated.
 A ground-level leakage of about 1 sqm should be postulated. The resulting fission
product release rate strongly depends on successful prior containment venting.
Core-concrete-interaction in cavity filled with water
 In case that the cavity is water flooded, a continous coolability can be realised and
the erosion process can be stopped if the so called EPRI-criterion for long time
debris coolability is fulfilled. An uncertainty distribution is suggested around this
criterion.
 In case of water flooded cavity without long time coolability, erosion rate and gas
release should be handled in the same way as in the case with dry cavity, taking into
account a steam source relevant to containment pressure.
Core-concrete-interaction
Phenomenon
or Process
Coolability of
core debris and
arrest of core
concrete
interaction
Recommendation
Rationale
Dependent on the particle size distribution:
5% percentile: 0.05 MW/m2
50% percentile: 0.2 MW/m2
95% percentile: 1 MW/m2
A large set of
experiments are
available in the
literature
Core-concrete-interaction
Phenomenon
Recommendation
Rationale
or Process
Pressurisation of Plant specific calculations necessary to assess
the containment the time period available for counter measures,
e.g. venting
 pressure build-up is caused by
- heat rate (residual heat generation is unknown),
- ratio of steam production and condensation (heat sinks) as well as
- produced non-condensable gases (MCCI)
 containment failure pressure is type specific
Pressurisation of Containment
 The extension of probabilistic analyses in the frame of PSR regarding severe
accident sequences with core melt as well as all requirements regarding extent,
methods and procedural steps are oriented on basic international recommendations
on procedures for conducting level 2 PSA, published e.g. in /IAEA 95/ or /NEA 97/.
 Detailed methodological requirements based on plant-specific German experiences
and current knowledge about core melt phenomena according to the state of the art
complete the level 2-part of the guidance.
 A systematical approach is recommended to derive reasonable conditional probability
ranges for these phenomena and processes.
 Focus is on those phenomena that are of importance with respect to their
consequences and to which a large uncertainty is associated. As far as possible
quantitative values are introduced.
Conclusions