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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
RELKO Ltd. Engineering and Consulting Services
FULL POWER AND SHUTDOWN LEVEL 2 PSA STUDY FOR
UNIT 1
OF J. BOHUNICE V1 NPP
by
Zoltan Kovacs and Helena Novakova
RELKO Ltd, Engineering and Consulting Services
Bratislava, Slovak Republic
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CONTENTS

Introduction and overview of the Level 2 PSA methodology

Description of the Confinement

The interface between the level 1 and 2 PSA

Description of the accident progression analyses

Evaluation of the confinement failure modes and
confinement event trees

Definition of release categories

Discussion of the source term analysis

The results and sensitivity studies

Conclusions
construction of the
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INTRODUCTION


Within the gradual reconstruction of J. Bohunice V1 NPP the upgrading
of the confinement was performed and new accident localisation system
was installed.
The level 2 PSA (for full power, low power and shutdown operating
modes) was developed with the following objectives:
 to identify the ways in which radioactive releases from the plant can occur
following the core damage,
 to calculate the magnitudes and frequency of the release,
 to provide insights into the plant behaviour during a severe accident,
 to provide a framework for understanding containment failure modes, the impact
of the phenomena that could occur during and following core damage and have
the potential to challenge the integrity of the confinement,
 to support the severe accident management and development of SAMGs.
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INTRODUCTION




The level 2 PSA model of the J. Bohunice V1 NPP was developed in the
RISK SPECTRUM Professional code.
This model calculates the frequency of the individual release categories
generating minimal cut sets which involve the initiating event of the
accident, component failures and human errors.
The magnitudes of release categories are calculated using: the
MAAP4/VVER for reactor operation and shutdown mode with closed
reactor vessel and the MELCOR code for shutdown mode with open
reactor vessel.
Although the level 1 PSA mission time is 24 h, the level 2 PSA simulated
accident sequences 48 h to provide greater understanding of confinement
performance during the later stages of a scenario. So, both the level 2
PSA mission time and the deterministic analyses time are 48 h.
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OVERVIEW OF THE LEVEL 2 PSA METHODOLOGY




Project management:Definition of objectives and scope of level 2 PSA
study, project management, team selection and organisation, quality
assurance of the project.
Familiarisation with the plant: Familiarisation with the plant and
identification of design aspects importance to severe accidents,
description of the confinement and accident localisation system of the
plant after the safety upgrading.
Interfacing of level 1 and 2 PSA: Development of extended event trees,
definition of plant damage states as initiating events for CETs.
Accident progression analyses: Analyses of progression of severe
accidents, computer codes used for the analyses, treatment of the accident
phenomena, input data, calculation results.
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OVERVIEW OF THE LEVEL 2 PSA METHODOLOGY





Confinement performance analyses: Structural response, confinement
bypass and confinement isolation analyses.
Construction of CETs: Construction of confinement event trees,
quantification of confinement event tree events, uncertainties in the event
probability quantification.
Source term analyses: Definition of release categories (sources terms),
grouping of fission products, fission product release calculations,
treatment of uncertainties in the estimated source terms.
Quantification of frequencies for release categories: Calculation of
frequencies of release categories using the integrated full power and
shutdown model developed in RISK SPECTRUM Professional code.
Presentation and interpretation of the results
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OVERVIEW OF THE LEVEL 2 PSA METHODOLOGY main events of the severe accident
Core
Damage In-Vessel Phase
(PDS)
Radioactive
release
(RC)
Reactor
Vessel
Melt-through
Late Ex-Vessel Phase
Intact
Confinement
Radioactive
release
(RC)
Radioactive
release
(RC)
Radioactive
release
(RC)
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CONFINEMENT DESCRIPTION
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CONFINEMENT DESCRIPTION - the main issues of the
confinement






The confinement leak-tightness
The accident localisation system
The confinement spray system
The confinement isolation
The confinement data as input to the deterministic codes
Structural analyses of the confinement
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CONFINEMENT DESCRIPTION
Leakage from the confinement: volume % / 24 h
Unit 1
Unit 2
600
L e ak ag e fr o m
Year
co n fn e m e n t
[ % / 24 h o u r s ]
U n it 1
U n it 2
1990
5039.00
7173.00
1993
403.96
565.00
1994
255.06
291.30
1995
132.00
93.00
1996
86.90
70.68
1997
68.18
51.04
1998
59.58
45.8
1999
58.16
43,16
2000
54.56
42.14
2001
48.5
40.77
500
400
300
200
100
U nit
0
1993
1994
1995
1996
1997
1998
1999
2000
2001
YEA R
U nit
2
1
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CONFINEMENT DESCRIPTION
R302
R301
p27kPa
12 x DN1130
DN1200
DN600
p 50kPa
R104
p15kPa
R102
p 30kPa
4xDN800
R002
2xDN250
1kPa
2xDN510
R048
ECCS
4xDN510
BWST
SS
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THE INTERFACE OF LEVEL 1 AND 2 PSA






Development of the level 2 probabilistic model is started with the
construction of the extended event trees (EETs).
This is complementary level 1 modelling before the plant damage state
grouping (PDSs).
It allows credit for the core damage recovery. Construction of EET is
performed for each core damage sequences of the level 1 PSA model.
The next step is a definition of PDSs and assignment to consequences of
the EETs.
Then, the confinement event tree (CET) is developed for each PDS as
part of level 2 probabilistic model.
Consequences of the CETs are the release categories. Their frequency
represents the results of the level 2 PSA.
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THE INTERFACE OF LEVEL 1 AND 2 PSA
INITIATING
EVENTS
LEVEL 1 EVENT TREES
CONFINEMENT EVENT TREES
(CET)
RELEASE CATEGORIES
(RC)
OK
CD
IE1
IE2
IE3
.
.
.
.
.
.
.
.
IEN
EXTENDED
LEVEL 1 EVENT TREES
-Nominal leakage
-Scrubber release
-Confinement rupture
-Basemat melt through
-Core damage recovery
(CDR) with leakage
OK
CDR
1
2
3
.
.
.
.
.
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THE INTERFACE OF LEVEL 1 AND 2 PSA



In same cases the level 1 event trees are changed before starting of
EET construction. The reason is in the definition of the event tree top
events.
The following example clarifies the problem. Given loss of the
primary to secondary side heat removal, failure of the primary bleed
and feed (top event in the level 1 event trees) via the pressurizer safety
valve leads to the core damage.
The reason of a failure (operator error or hardware failure) is
evaluated in the fault tree under the top event. However, different
plant damage states occurs if operator fails to initiate bleed or bleed
and feed is initiated but the losses are not compensated by at least a
HPSI pump.
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THE INTERFACE OF LEVEL 1 AND 2 PSA



Development of the level 2 model of the Bohunice V1 plant was
performed using the RISK SPECTRUM PSA Professional code. The
level 1 PSA model was developed also in this code.
After an evaluation of the software capabilities for level 2
applications, it was decided to streamline the sequence quantification
process by: 1) adding the EET top events directly to the core
damage
sequences of the level 1 event trees,
2) defining the PDSs as the initiating events of the CETs and
3) assign the release categories to the CET sequences.
In principle, this allows the explicit consideration also of all level 1
events in the overall model. The level 2 calculation may therefore be
based on the complete plant model without any intermediate
results.
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THE INTERFACE OF LEVEL 1 AND 2 PSA

The top events of the extended event trees:





confinement isolation
HPSI
LPSI
Confinement spray system
Other events
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THE INTERFACE OF LEVEL 1 AND 2 PSA: PDS


The plant damage states (PDS) represent a functional groupings of level
1 core damage sequences.
The criteria for binning the level 1 sequences into the plant damage states
are based on the following five characteristics of each sequence:
 Initiator (Large LOCA, Small LOCA, Transient, Confinement bypass)
 Time to core melt (early < 1 h, late > 1 h)
 ECCS status (A - water injected into the RPV or reactor cavity, core damage
recovery possible, D - no water injected)
 Confinement spray system (Y - available, N -unavailable)
 Confinement status (I - isolated, A - not isolated, B - bypassed)
Example of PDS: TLDNI - transient, late CD, no CD recovery, no spray
system, confinement isolated
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ACCIDENT PROGRESSION ANALYSES



The J. Bohunice V1 level 2 PSA is accomplished by coupling a probabilistic
model of the confinement response to the postulated initiating events with a
deterministic physical model to examine the plant response.
This process also incorporates the evaluation of the impact of the
phenomenological uncertainties.
The probabilistic model is embodied in EETs and CETs developed in the
RISK SPECTRUM PSA code. The plant physical model is defined in the
MAAP4/VVER and MELCOR parameter files (more detailed description
is in presentation of Mr. M. Cvan from VUJE).
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ACCIDENT PROGRESSION ANALYSES
• The deterministic codes provide information required to perform the
calculations of the plant specific fission product transport and thermal
hydraulic response to the postulated accident sequences.
• They are also used to study the sensitivity of the source term to the
phenomenological uncertainties.
• The source term analysis used a default values for the model
parameters. The sensitivity analysis identifies any variations from this
approach.
• The deterministic analyses are supplemented with the
phenomenological evaluation summaries to provide a complete
physical representation of the plant.
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CONFINEMENT FAILURE MODES
• Phenomenological evaluations have been performed in support of
the plant level 2 PSA project to determine the likelihood of all
postulated confinement failure modes and mechanisms.
• These evaluations were performed systematically to address the
controlling physical processes or events specific to the plant
configuration. The confinement failure modes of the plant considered
unlikely are: over-pressure, direct confinement heating, steam
explosions, thermal attack of the confinement penetrations, vessel
thrust forces and melt through the induced failure of the reactor cavity
floor.
• Failure modes more likely to occur are: hydrogen combustion,
confinement isolation failure, and confinement bypass.
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CONFINEMENT FAILURE MODES - Cofinement structural
analyses
• A plant specific structural analysis of the J. Bohunice V1
confinement (SG compartment) has been performed to
determine the ultimate internal temperature and pressure
capacity and the most likely failure locations for BDBA (break of
the cold leg of a RCS loop with double ended discharge of primary coolant;
no ECCS and no spray system is considered).
• The BDBA present the maximum temperature and pressure
load for the confinement. For all initiators within the DBA
lower values of these parameters were calculated by the
MAAP4/VVER code.
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CONFINEMENT FAILURE MODES - Cofinement structural
analyses
- Temperature loading:
The temperature in the hermetic zone did not increase during the LOCA 160°C in
accordance to MAAP4/VVER analysis.
The effect of this temperature on the confinement integrity is minimal during the
accident.
The results of analyses of the concrete structures under the accident temperature
loads present that the structure resistance is sufficient for the BDBA loads condition.
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CONFINEMENT FAILURE MODES - Cofinement structural
analyses
- Pressure loading:
The structural analysis carried out for BDBA under the high internal
over-pressure corresponding to the mean strengths and non-linearity
behaviour of the concrete structures.
The evaluation of the structural integrity was performed for the critical
places, which were defined from the previous non-linear analysis for
various loads of BDBA and DBA.
The non-linear analysis takes into account the concrete cracking and
crushing, layered approximation of the shell elements with various
material properties, etc.
The maximum pressure in the confinement in case of large LOCA is 223
kPa. Probability of loss of confinement integrity for this pressure is less
than 0.0001.
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CONFINEMENT FAILURE MODES -unlikely failure modes



Direct confinement heating (DCH) is the process of directly heating the
confinement atmosphere by the molten core debris should it be hydrodynamically forced out of the reactor cavity due to the primary system
blowdown. A phenomenological evaluation was performed to examine the
likelihood of the plant confinement failure due to DCH. The evaluation excluded
this event.
Thermal attack of confinement penetrations. The mechanical and electrical
penetrations or seals are not susceptible to thermal degradation due to the
confinement gas temperatures. Those penetrations can withstand the confinement
temperatures up to and beyond 500 C. However, such elevated temperatures are
not predicted for the confinement for sustained periods of time.
Ex-vessel steam explosion. Unlikely event because the reactor cavity is dry.
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CONFINEMENT FAILURE MODES - unlikely failure modes

Vessel thrust force: The maximum jet thrust force could not lift the
vessel and its internals, even without considering the ability of the vessel
support structure to withstand the thrust load.
If the coolant loop piping and shield wall are considered, a much larger
force would be required to dislodge the reactor vessel. Even if the vessel
could shirt, the confinement is configured in such a manner that the
reaction forces cannot be transmitted to the confinement wall.
Therefore, this postulated failure mode is prevented by the plant design.
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CONFINEMENT FAILURE MODES - unlikely failure modes

Confinement failure by melt through: If the basemat melt through
occurs, the atmospheric release will be limited to that resulting from the
design-basis leakage and a filtered release of radionuclides through the
soil.
Because of the presence of the water on the confinement floor, it was our
judgement that this event is precluded. However, if all water pathways
are plugged, a melt through could eventually occur after tens of hours. A
probability of 10- 4 with an error factor of 10 was assigned to this event
for all PDS and sequences on the basis of engineering judgement. Later it
was removed from the CETs because the risk associated with this event
would not affect the total risk.
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CONFINEMENT FAILURE MODES - Hydrogen detonation
(considered event)



From the analyses it was considered that the main contributor to the
failure of the confinement is from the ignition of the combustible gas
mixtures.
A best-estimate assessment of the in-vessel and ex-vessel hydrogen
production is possible using the MAAP4/VVER code. The code
calculates the hydrogen, oxygen and steam inventory in the confinement
and identifies the time periods when the hydrogen is combustible or
invent.
The potential confinement pressurization resulting from the hydrogen
combustion is bounded by the calculating the adiabatic isochoric
complete combustion (AICC) of this assumed hydrogen inventory. It is
hand calculated on the basis of the outputs from the MAAP4/VVER
code.
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CONFINEMENT FAILURE MODES - Hydrogen detonation
(considered event)
• The analyses identified three phases of possible hydrogen burn:
before the reactor vessel failure, at the reactor vessel failure and
after the reactor vessel failure.
• However, hydrogen detonation is precluded if the confinement
is not isolated.
• In the level 2 PSA of western plants the likelihood
of hydrogen detonation is set 0 for PDS with the confinement spray
system fails to operate for reduce the steam concentrations.
However, for the J. Bohunice V1 plant the MAAP4/VVER
analyses have shown that the hydrogen detonation is possible
also if the spray system fails to operate.
• This hydrogen detonation concentration exists
in the BWST which is part of the confinement boundary
(the steam is condensated but the hydrogen is accumulated above the
water level). In other areas of the confinement (in the SG boxes)
the hydrogen detonation is not possible if the spray system fails to operate.
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CONFINEMENT FAILURE MODES - Confinement isolation
failure (considered event)



The confinement isolation system is designed to preserve the ability of
the confinement boundary to prevent or limit the escape of fission
products that may result from postulated accidents.
In the event of a possible radiation release from the confinement through
the process lines, the confinement isolation system automatically isolates
all lines penetrating the confinement which do not serve an accident
mitigating function.
For the Bohunice V1 plant the confinement ventilation system and the
BWST blow- down line must be isolated in case of the accident. Fault
tree was developed and involved into the level 2 PSA model.
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CONFINEMENT FAILURE MODES - Confinement bypass
(considered event)



Confinement bypass is considered as an accident initiator that can lead to
the core damage because the loss of cooling fluid to a location outside the
confinement disables the ECCS for long-term core cooling.
The most likely mechanisms for this failure mode, identified for the plant
as being significant in terms of the potential consequences, are SG
collector rupture (SGTM), SG tube rupture (SGTR) or interfacing
LOCA.
Note, however, that the SGTM sequences during the full power operation
contribute about 36% to the total core damage frequency. Contribution to
the full power core damage frequency from SGTR is 3.4% and from
interfacing LOCA is less than 1%. These initiators are not dominant from
the risk point of view during the reactor shutdown.
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CETs
The general guidelines used for the development of the CETs are
summarised below:
- the initiating event of a CET is a PDS,
- the CET top events and structure provide the details necessary to
characterise the fission product source term releases,
- the CET considers factors which dominate the confinement response;
thus, the top events consider broad categories of the confinement
behaviour,
- the CET considers early confinement failure timing (i.e., confinement
failure at or shortly after vessel failure) and late confinement failure; the
results indicate significant impact from early and late hydrogen
detonation.
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CETs - Main assumptions

The following main assumptions are used in the CET construction:
 .For core damage recovery only the HPSI and LPSI pumps of unit 1 are
considered; HPSI pumps of unit 2 are not taken into consideration due to
the limited water sources; external water sources are also not considered
for this purpose.
 It is possibility to supply the RCS also by the spray system pumps.
However, using a spray system pump for the core damage recovery is not
considered in the model because these pumps must perform other safety
functions.
 If the core damage recovery for PDS with early CD is not performed
before the vessel failure, recovery after the vessel failure is considered not
to be possible.
 .After successful core damage recovery no hydrogen detonation is
considered.
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CET: Large LOCA, early core damage, ECCS available, spray
system available, confinement isolated
Confinement event tree 1 LEAYI-A
CET1A
CONFINEMENT SPRAY AND
Core damage recovery
Hydrogen detonation at
COOLING OF THE TANK before reactor vessel failure reactor vessel failure
AVAILABLE
@SS(S&C)00
ECDR-1A
EHD
2
Hydrogen detonation after
reactor vessel failure
LHD
1
Conseq.
NR
2
2
RC1
1
3
RC4.2
4
RC3
2
5
RC2
1
6
RC4.2
7
RC3
1
2
1
No.
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RELEASE CATEGORIES

Four general classes of the containment-failure modes were involved into
the RC:
 Isolation failure
 Hydrogen detonation
 Design leakage
 Confinement bypass (SGTM, SGTR, interfacing LOCA)

These general classes represent different source term magnitudes because
they represent different leakage rates: leakage via not isolated piping,
gross structural failure, low gradual release via confinement normal
leakage and confinement bypass with different rates.
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RELEASE CATEGORIES - leakage rate




The blow-off line of 1 200 mm diameter (installed from the borated
water storage tank to the reactor hall) is the most dominant leakage path,
if it is not isolated.
Hydrogen detonation in the confinement is defined as increased leakage
with gross structural failure leading to a puff release of radionuclides
followed by leakage through an open path to the environment.
The design leakage involves releases via COFs and normal leakage of the
confinement (the leakage rate is 48 volume percent per day).
The confinement bypass occurs after SGTM, SGTR and interfacing
LOCA
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RELEASE CATEGORIES

In addition, the following issues were integrated into the RCs:
 Release mechanism (core overheating, in-vessel and ex-vessel core
damage recovery)
 Effects of the spray system operation
 Time to core damage and time to vessel failure
 Time to confinement failure
 Plant operating modes
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RELEASE CATEGORIES - times to core damage and vessel
failure
Initiating event
LOCA 500 mm
LOCA 100 mm
LOCA 32 mm
PDS
LEAYI, LEDNI, LEDNN,
LEDYN, LEDYI
SEAYI, SEDNI, SEDNN,
SEDYN, SEDYI
SLAYI, SLDNI, SLDNN,
SLDYN, SLDYI
TLAYI, TLDNI, TLDNN,
TLDYN, TLDYI
Loss of secondary
heat removal, no
bleed and feed
Loss of RHR, open TLDNN, TLDYN
reactor vessel
SLDNN, SLDYN
Time to core damage Time to vessel failure
[h]
[h]
0.22
1.80
0.45
3.00
2.18
5.18
4.20
6.00
6.90
21.50
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RELEASE CATEGORIES
 NR
 RC0.1
 RC0.2
 RC1
 RC2
 RC3
 RC4.1
 RC4.2
 RC5.1
 RC5.2
 RC6.1
 RC6.2
- no release
- confinement survives with spray, in-vessel core damage recovery
- confinement survives with spray, ex-vessel core damage recovery
- confinement survives with spray, no core damage recovery
- confinement survives with spray, no core damage recovery
- early confinement failure at vessel failure
- late confinement failure at vessel failure
- late confinement failure after vessel failure
- confinement isolation failure with spray, reactor vessel closed
- confinement isolation failure with spray, reactor vessel open
- confinement isolation failure without spray, reactor vessel closed
- confinement isolation failure without spray, reactor vessel open
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RELEASE CATEGORIES
 RC7.1 - confinement bypass after SGTM, steam dump station to the atmosphere
re-closed
 RC7.2 - confinement bypass after SGTM, steam dump station to the atmosphere
fails to re-close
 RC8.1 - confinement bypass after SGTR, steam dump station to the atmosphere
re-closed
 RC8.2 - confinement bypass after SGTR, steam dump station to the atmosphere
fails to re-close
 RC9 - confinement bypass after interfacing LOCA
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SOURCE TERMS CHARACTERISATION



The purpose of the source term analysis is to quantitatively describe the
magnitude and composition of radionuclide releases to the environment
resulting from the core damage accidents.
Before the source term calculations were actually performed, the
sequences with similar source term characteristics were grouped into the
release categories to reduce the total number of the sequences to be
analysed.
Source term quantification was then performed by the analysing a single,
representative accident sequence for each release category by the
MAAP4/ VVER code resp. MELCOR code.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
RELKO Ltd. Engineering and Consulting Services
SOURCE TERMS CHARACTERISATION - sequence
selection



A representative systemic sequence for each release category was
selected for the source term analysis.
The analysed sequence was chosen because it had the highest frequency
of occurrence of any sequence within the release category or because it
was expected to bound all other sequences of these category.
Selection of a sequence other than that with the highest frequency
occurred when that sequence could result in earlier core damage and
vessel failure or it was not a full power sequence. To be conservative,
always a full power sequence is selected for the plant operating modes
with the closed reactor vessel.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
RELKO Ltd. Engineering and Consulting Services
RESULTS - power operation - conditional probabilities of RC
Release
category
RC7.1
RC4.1
NR
Frequency
RC2
RC6.1
1.38E-06
1.36E-06
RC3
1.15E-06
RC8.1
RC0.2
9.59E-07
6.61E-07
RC9
2.73E-07
RC4.2
1.46E-07
RC5.1
RC0.1
1.15E-07
6.78E-09
RC1
RC8.2
8.65E-11
7.42E-11
RC7.2
1.24E-11
8.39E-06
6.05E-06
6.03E-06
Definition
Conditional
probability
Confinement bypass after SGTM, > 10% volatiles released
0.30
Late confinement failure at vessel failure, > 10% volatiles released
0.22
Success, no confinement failure within 48 h, < 0.1% volatiles
0.22
released
Confinement survives with sprays, > 10% volatiles released
0.05
Confinement isolation failure without spray, > 10% volatiles
0.05
released
Early confinement failure at vessel failure, > 10% volatiles
0.04
released
Confinement bypass after SGTR, > 10% volatiles released
0.04
Confinement survives without spray, late core damage recovery,
0.03
1% volatiles released
Confinement bypass after interfacing LOCA, > 10% volatiles
0.01
released
Late confinement failure after vessel failure, > 10% volatiles
0.01
released
Confinement isolation failure with spray, 4 volatiles released
0.01
Confinement survives with spray, early core damage recovery,
0
< 1% volatiles released
Confinement survives with spray, 5% volatiles released
0
Confinement bypass after SGTR, stuck open steam dump station
0
to the atmosphere, > 10% volatiles released
Confinement bypass after SGTM, stuck open steam dump station
0
to the atmosphere, > 10% volatiles released
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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RESULTS - power operation - LERF



The large early release frequency is calculated from the release
categories, where more than 10% volatiles is released and the release is
initiated within 2 h after initiating events. So, the large early release
frequency (LERF) is given as a sum of the following frequencies:
LERF = RC3 + RC5.1 + RC6.1 + RC7.1 + RC7.2 + RC8.1 + RC8.2 +
RC9 = 1.22E-5/y
Conditional probability of large early release given core damage is 0.44.
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RESULTS - shutdown
POS/RC
POS1
POS2
POS3
POS4
POS5L
POS5S
POS6
POS7
POS8
POS9
POS10
SHUTDOWN
Total
NR
6.10E-08
6.10E-08
RC0.1
8.56E-12
2.20E-11
6.68E-11
1.76E-10
RC0.2
1.95E-09
1.15E-09
3.91E-08
3.70E-09
2.79E-08
7.32E-08
RC1
2.38E-13
6.76E-13
1.26E-12
3.19E-13
1.62E-12
4.03E-12
RC2
1.41E-09
9.89E-10
1.57E-09
2.72E-09
7.48E-09
1.42E-08
RC3
2.09E-09
4.13E-09
1.26E-08
1.88E-08
RC4.1
1.71E-08
1.04E-08
3.52E-07
3.27E-08
2.54E-07
6.66E-07
RC4.2
2.29E-10
2.41E-11
7.36E-10
4.59E-10
1.78E-09
3.23E-09
POS/RC
POS1
POS2
POS3
POS4
POS5L
POS5S
POS6
POS7
POS8
POS9
POS10
SHUTDOWN
Total
RC5.1
6.03E-11
3.08E-11
3.45E-06
6.91E-06
1.27E-09
1.42E-10
6.32E-10
1.04E-05
RC5.2
1.17E-05
1.36E-06
1.57E-05
2.88E-05
RC6.1
6.37E-10
4.66E-10
2.86E-07
5.52E-07
9.29E-10
2.82E-09
1.16E-08
8.54E-07
RC6.2
5.77E-06
1.31E-06
2.80E-06
6.76E-06
1.67E-05
RC7.1
1.90E-08
1.38E-08
2.13E-08
3.60E-08
9.55E-08
1.86E-07
RC7.2
2.78E-14
2.03E-14
3.11E-14
5.22E-14
1.38E-13
2.66E-13
RC8.1
1.60E-09
1.08E-09
1.89E-09
3.16E-09
9.51E-09
1.72E-08
RC8.2
1.72E-13
1.26E-13
1.92E-13
3.22E-13
8.68E-13
1.68E-12
RC9
5.66E-10
3.78E-10
5.92E-10
1.08E-09
2.35E-09
3.73E-09
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
RELKO Ltd. Engineering and Consulting Services
RESULTS - shutdown - LERF

So, the large early release frequency (LERF) is given as a sum of the
following frequencies:
LERF = RC3 + RC5.1 + RC6.1 + RC7.1 + RC7.2 + RC8.1 + RC8.2 +
RC9 = 1.15E-5/y

Conditional probability of large early release given core damage is 0.21.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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Sensitivity analyses




Sensitivity analyses were performed to address the factors which can
have impact on the release category frequency and magnitude. Sensitivity
analyses were also performed to address the questions concerning
equipment operation during a severe accident.
The factors:
- Large confinement bypass (recovery action)
- Hydrogen recombiners
After implementation of changes the large early release frequency will be
3.75E-6/y for power operation.
No impact of changes on the shutdown risk, due to the high CDF
(symptom based procedures have to be implemented)
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
RELKO Ltd. Engineering and Consulting Services
Sensitivity analyses -equipment survivability inside
confinement

After reviewing the environmental parameters of a severe accident and
the critical components located inside the confinement, it is concluded
that the issues associated with the equipment operability inside the
confinement are: aerosol accumulation on the spray nozzles and high
temperature, plugging and radiation.

The engineered safety features are expected to survive the pressure,
temperature, radiation, debris and steam conditions expected during a
severe accident. This evaluation covers critical equipment located inside
the confinement and the safety equipment building.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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Sensitivity analyses -equipment survivability outside
confinement



If the HPSI, LPSI pumps and confinement spray pumps are not lost in a
severe accident, then the critical components of these pumps can be
adequately cooled and maintain operability in the recirculation mode.
This conclusion is based on a review of the BWST water temperature,
temperature of critical pump components and the pump motor cooling.
In addition, a review of environmental qualification results of the pump
power cables indicated that these cables will remain operable at the
elevated room temperature. A pump room equipment inventory survey
shows that there is no other heat sensitive equipment inside the room.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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CONCLUSIONS

For the full power operation it can be concluded that:
 The results indicate that, given core damage, there is an 25% probability that the
confinement will successfully maintain its integrity and prevent an uncontrolled
fission product release. After the implementation of the recovery actions for
SGTM and installation of hydrogen recombiners in the confinement this
probability will be increased to 74%. For comparison: a western PWR plant has
the probability of 84%.
 The most likely mode of release from the confinement is a confinement bypass
after SGTM with conditional probability of 30%. Late confinement failure (after
6 h) at the vessel failure, with a conditional probability of 22%, is the next most
likely mode of the fission product release. Finally, the confinement survives with
the spray is expected to occur with a conditional probability of 5% per core
damage event. The conditional probability for the confinement isolation failure
probability without spray is 5%, for early confinement failure at the vessel
failure is 4%, for other categories 1% or less.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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CONCLUSIONS
 .The overall conditional confinement failure probability of 75% by the proposed
modifications can be decreased to 26%. For a western plant this value is 16%.
 The results of the level 2 PSA indicate that there are vulnerabilities in the area of
the protection against hydrogen detonation. It requires immediate attention to
improve the plant risk profile. In addition, attention must be paid to development
of SAMGs in coincidence with the conclusions of this study.
 Vulnerability screening was performed based on the screening criteria provided
for US plants in “Criteria for Selecting Important Severe Accident Sequences”.
The criteria states as follows: “Any functional sequence that has a core damage
frequency greater than or equal to 1.0E-6 per year and that leads to containment
failure which results in a radioactive release magnitude greater than or equal to
PWR-4 release category of WASH-1400”. The PWR-4 release category was
estimated as 10% of the volatile fission products. For the full power operation
two such release sequences exits for category RC4.1, one sequence for RC6.1
and one sequence for RC7.1. However, the proposed modifications will remove
the sequences from the list of the important sequences in case of RC4.1 and
RC7.1.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
RELKO Ltd. Engineering and Consulting Services
CONCLUSIONS

For the shutdown operating modes it can be concluded that:
 The shutdown risk is high for the open reactor vessel and open confinement. The
reason is the high core damage frequency in the shutdown operating modes.
After implementation of the recommended shutdown symptom-based emergency
procedures significant decrease of the shutdown risk is possible.
 Installation of filtered venting system in the reactor hall with long term operation
could significantly decrease the release magnitudes during the shutdown
operating modes.
 .Important severe accident sequences based on the above mentioned definition
exists for release categories: RC5.1, RC5.2 and RC6.2.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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CONCLUSIONS

A number of features were identified through the course of the level 2
PSA which contribute to the performance of the confinement:
 The most important feature of the confinement with respect to the fission product
retention is its ability to remain intact in case of the steam over-pressurisation.
This construction allows natural deposition mechanisms to remove the airborne
fission products from the confinement atmosphere, and provides adequate time
for the additional accident mitigation activities to be implemented.
 .Installation of hydrogen recombiners are extremely important for the severe
accident conditions. The confinement can not withstand hydrogen detonation.
 .The inability to get the water into the reactor cavity prevents the external
cooling of the intact reactor vessel.
 The absence of any penetration in the lower vessel head coupled with the natural
circulation in the primary system during a high pressure core melt is expected to
induce creep rupture failure in the hot leg pipe prior to the vessel failure.
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International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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CONCLUSIONS
 .Depressurization of the primary system prior to the vessel failure, as a result of
the creep rupture failure of a hot leg pipe, should preclude concerns about the
high pressure severe accident phenomena (i.e., ex-vessel steam explosion, direct
confinement heating and vessel thrust forces).
 Retention of the core debris in a dry cavity may induce MCCI melt through the
cavity floor if the core debris cannot be cooled by the water.
 Injecting water through the failed reactor vessel, in an attempt to cool the core
debris in the cavity, is advisable and it does not depend on the status of the
confinement cooling. Injecting water into a cavity filled with the hot core debris
results in the formation of the hot steam. If the confinement cooling is available
this steam is condensed. If no confinement cooling is available steaming from
the cavity can eventually over-pressurise the confinement but the COFs prevent
the confinement failure. The radiological release consequences of inducing a
COF opening are expected to be lower than those associated with MCCI failure
of the cavity floor.
53/54
International Workshop on Level 2 PSA and Severe Accident Management, 29-31.3.2004 Koln, Germany
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CONCLUSIONS
 Steam inerting of the confinement during a severe accident can not prevent
hydrogen detonation in the BWST.
 The insights gained through the analysis of the severe accident progression and
the detailed study of the related phenomena has provided a detailed
understanding of the plant behaviour under the severe accident conditions. The
knowledge developed can form the basis for the future developments of SAMGs.
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