Post on 14-Jun-2020
National Commission for Nuclear Activities Control
ROMANIA
Re-assessment of Cernavoda nuclear power plant
design safety in the aftermath of the Fukushima
Daiichi accident
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi
Accident, Vienna, Austria, from 26 to 29 August 2013
Cantemir Ciurea - Ercau, Director, Nuclear Fuel Cycle Division
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Cernavoda NPP
Reactor Type Gross Capacity
MW(e)
Construction
Start First Criticality Status
Cernavoda-1 CANDU-6 706.5 1980 16th of April
1996 In operation
Cernavoda-2 CANDU-6 706.5 1980 6th of May 2007 In Operation
Cernavoda-3 CANDU-6 720 1980 - Under
Preservation
Cernavoda-4 CANDU-6 720 1980 - Under
Preservation
Cernavoda-5 CANDU-6 - 1980 - Under
Preservation
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Safety re-assessments of Cernavoda NPP
after Fukushima NPP accident
Romania has implemented actions in line with the initiatives taken at
international level:
The first actions taken by the operator were based on the recommendations
in the Significant Operating Experience Report issued by the World
Association of Nuclear Operators (WANO), WANO SOER 2011-2,
“Fukushima Daiichi Nuclear Station Fuel Damage Caused by Earthquake and
Tsunami”.
The actions devised by CNCAN were based on the specifications for “stress
tests” required by the European Commission for all the nuclear power plants
in the European Union, based on a proposal developed by WENRA (Western
European Nuclear Regulators Association) and agreed by ENSREG
(European Nuclear Safety Regulators Group).
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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The “stress test” safety reassessment consisted of:
• An evaluation of the response of a NPP when facing a set of extreme situations (e.g.
beyond design basis earthquake, beyond design basis flood, severe weather, etc.);
• A verification of the preventive and mitigative measures chosen following defence-
in-depth logic: initiating events, consequential loss of safety functions, severe
accident management
The following topics were taken into account :
• Topic 1: Initiating events (external events)
• Topic 2: Loss of safety functions
• Topic 3: Severe Accidents Management
In this assessment it was required to consider that the plant has to use only the
on-site resources (conventional and nonconventional equipment, fuel, oil, water,
etc.) for at least 72 hours.
The results of the “stress test” assessment have been subject to a peer-review
and several recommendations were received, in addition to the improvements
already identified and under implementation.
Safety re-assessments of Cernavoda NPP
after Fukushima NPP accident
External Events
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
External Events
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The following aspects were addressed during the review of design
provisions against seismic, flooding and extreme weather
scenarios:
- Design basis and compliance with the design basis;
- Assessment of plant behaviour for beyond design basis external
events, quantification of safety margins and identification of any
potential cliff-edge effects;
- Identification of measures to increase the robustness of the plant for
extreme external events;
- Planned actions to improve safety.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Design Basis Earthquake
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According to the seismic qualification principles adopted for CANDU
6 NPPs, the design considers two seismic activity levels, both
imposed by nuclear safety requirements, namely:
Design Basis Earthquake (DBE) level - the engineering representation
of earthquakes generating the worst possible severe effects, applicable
to a NPP site and having a sufficiently low probability of being exceeded
during the plant lifetime.
For Cernavoda NPP, the DBE has a return frequency of 1E-03
events/year and based on the site specific hazard curve the intensity
and acceleration are I = VIII degrees MSK-64 and Peak Ground
Acceleration (PGA) = 0.2 g.
Site Design Earthquake (SDE) level - represents the seismic activity
with the period of return for the NPP site 100 years. This is the seismic
design level for the plant systems that must remain operational for a
long period of time after a loss of coolant accident (LOCA).
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Validity of Data in Time & Conclusion on the
Adequacy of the Design Basis
In 2004 an update of probabilistic seismic hazard analysis (PSHA)
was carried out for the Cernavoda NPP site.
That PSHA was carried out to support a seismic Probabilistic Safety
Assessment (PSA) for the existing reactors at the Site.
The main conclusions from this study are:
The Vrancea sub-crustal seismic source dominates the seismic hazard at
the Cernavoda NPP.
For the maximum historical recorded event, with a magnitude M = 7.5 the
corresponding PGA at the NPP rock surface is 0.11g.
For the maximum estimated event with a magnitude M = 7.80, the PGA at
the NPP rock surface is 0.18g.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Provisions to protect the plants against the Design
Basis Earthquake
CANDU 6 reactors are designed for safety with a philosophy to deal with
design basis accidents (DBA) and DBE events with significant margins. Both
diverse and redundant systems are implemented to ensure safe reactor
shutdown and fuel integrity.
The safety-related systems, structures and components (SSCs) are divided
into two groups as follows:
Group 1: PHT, shutdown system one (SDS1), main and auxiliary
moderator, steam and feedwater system, emergency core cooling (ECC)
system, shutdown cooling (SDC), Local Air Coolers (LACs), Class I, II, III
and IV power, and the main control room (MCR).
Group 2: shutdown system two (SDS2), Containment Structure,
Containment Isolation System, Dousing, Air locks, Hydrogen Control, EWS,
EPS and the secondary control area (SCA).
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Evaluation of seismic margin
The seismic margin analyses have been based on the well established
methodology and on the reports elaborated as part of the Seismic PSA
performed in the period 2004 - 2005 for both Cernavoda NPP Units.
A review level earthquake (RLE) was established at a reasonably high
level seismic ground motion, based on site seismicity and plant specific
design features. The selected RLE has a return period of less than
1/10000 years, with a Ground Motion Response Spectrum (GMRS) with
a PGA of 0.33g.
Based on a review of the DBE qualified systems required for performing
the safety functions, a complete Safe Shutdown Equipment List (SSEL)
has been compiled.
In the framework of the stress test the FRS were pro-rated from 0.33
ZPGA for the targeted value of 0.4 g. to assess the seismic capacity of
SSC. The return frequency for this PGA is 5*E-05.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Conclusions and improvement measures
The plant seismic capacity was assessed for 0.4g (RLE 0.33g).
Up to this value, all the fundamental safety functions can be maintained.
The seismic margin assessment has not revealed a need for any safety
significant design change.
Several recommendations resulted from seismic walkdown, which have
been considered by the licensee as part of the regular plant seismic
housekeeping program.
The robustness of I&C panels has been increased to reduce the relay
chattering effects – (implemented on both units);
The seismic robustness of masonry walls in the CLI/II batteries room
has been increased (currently implemented on both units).
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
The “stress tests” required also an analysis of cliff-edge effects. For Cernavoda
NPP there are no cliff-edge effects occurring for PGA ≤ 0.4g. Therefore,
assessments of plant behaviour for PGA values greater than 0.4g have not
been performed, meaning that any additional seismic capacity above this value
has not been quantified.
Currently there is no agreed methodology for the performance of assessments
focused on cliff-edge effects rather than of seismic margins expressed in
HCLPF values. Further work will be done in this area, as recommended by the
peer-review, once an internationally agreed methodology becomes available.
Based on peer review recommendations, the adoption of a seismic level
comparable to the SL-1 of IAEA, leading to plant shutdown and inspection, is
currently under consideration.
Currently the actions taken by the licensee following an earthquake are based
on decision-making criteria that include the estimated damage to the plant
(walkdowns using a specific procedure) rather than on pre-defined ground
motion design response spectra.
Recommendations from peer-review
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Reassessment of plant design against flooding events
The assessment started with a systematic review of the original analyses
done for Cernavoda NPP at the siting stage in order to reassess the flooding
level that the plant can withstand without losing any of the main safety
functions that could lead to a severe accident event.
Cernavoda NPP site is located adjacent to the Danube River that is
providing required cooling water flow. The site is 60 km away from the Black
Sea coast.
The site is bordered on the Northeast by Cismelei Valley and on the
Southeast by a bypass channel of Danube-Black Sea Channel (DBSC).
Cooling water for the plant is taken from DBSC through a Bypass Channel,
Intake Channel and Distribution Basin.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Cernavoda site elevations
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Cernavoda Area – digital model.
Cismelei Valley.
Cernavoda Area
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
At the time of the selection of Cernavoda site, it was assumed that
two future dams would be built on the Danube River, one upstream of
Cernavoda and one downstream. These dams were never build but
the supporting studies carried out at that time analysed the different
regimes to determine the maximum (flood) water level of the dam
accumulation lake, and the extreme case of the upstream dam
breaking while the downstream dam holds.
Based on the original study, the maximum design water level for
return period of 1 in 10000 years for Cernavoda NPP is +14.13
mBSL.
The elevation of +16.00 mBSL for Cernavoda NPP site was selected
assuming the extreme postulated failure mode of the planned dams.
Flooding against which the plant is designed
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Reassessment of the Design Basis Flood
Potential external flooding sources considered by design for
Cernavoda NPP site are the following:
Extreme Danube River water flow / level;
Extreme local rainfalls.
The original DBF calculation has been reconfirmed by more recent
data and studies, the latest one having been completed in
September 2011, using the modern tool of Digital Topographic
Model (DTM) to create the external flooding hazard map for
Cernavoda NPP site and adjacent area.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Reassessment of the design basis flood has been made considering
the following events:
Extreme Danube River water flow / level;
Flooding due to rainfall on the Cernavoda site platform;
Flooding due to rainfall on catchment area;
Tsunami induced flooding;
Hydro-plant dam failure.
Reassessment of the design basis flood
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013 19
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013 20
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013 21
Worst-case scenario:
Break B4 location in-between Unit 1 and Unit 2
is the worst sensitivity case. The results show
that flowing water could increase up to 1 m
around 110 KV Station and up to 0.5 m behind
the Turbine Buildings were the main output and
unit service transformers are located.
It was conservatively assumed that this scenario
leads to initiating event Total Loss of Class IV power
(loss of off-site power), bounded by the potential
unavailability of Class III Standby Diesel Generators.
Plant response is provided by the existing abnormal
operating procedures at both Unit 1 and Unit 2, since
the buildings housing SSCs required to perform
essential safety functions (EWS/BMW and SCA/EPS)
are not subject to flooding.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Based on the analysis results obtained by making use of the latest
deterministic tools and complemented by probabilistic approach, it
was concluded that the Cernavoda NPP design intent in relation with
flooding hazards provides sufficient safety margins, therefore no
further measures for improvement were envisaged in this area.
Nevertheless, several measures to improve protection against
flooding by flood resistant doors and penetrations sealing have been
implemented for safety related equipment located in rooms below
plant platform level (such as the EPS, SCA, Service building, building
containing the SDGs fuel transfer pumps in Unit1)”. Also, sand bags
have been provided on-site to be used as temporary flood barriers, if
required.
Conclusions and improvement measures
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Extreme Weather Conditions
In order to derive a comprehensive list of severe weather events to be considered for
Cernavoda Units 1 and 2, a review of CNSC regulatory documents and guides, IAEA
guides, ANSI/ANS standards, and US NRC documents and IPEEE experience has been
performed.
The preliminary screening criteria, established in accordance with internationally recognised
standards, consist of the following (any one of these criteria was considered to be sufficient to
screen out the event):
Criterion 1: The event is of equal or lesser damage potential than the events for which the
plant has been designed. This requires an evaluation of plant design bases in order to
estimate the resistance of plant structures and systems to a particular external event.
Criterion 2: The event has a significantly lower mean frequency of occurrence than another
event taking into account the uncertainties in the estimates of both frequencies. The event in
question could not result in worse consequences than the consequences from the other
event.
Criterion 3: The event cannot occur close enough to the plant to affect it. This criterion must
be applied taking into account the range of magnitudes of the event for the recurrence
frequencies of interest.
Criterion 4: The event is included in the definition of another event.
Criterion 5: The event is slow in developing and it can be demonstrated that there is sufficient
time to eliminate the source of the threat or to provide an adequate response.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Extreme Weather Conditions
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Events screened out based on criterion 5 (slow developing event) have been
addressed in the national stress test report.
Low winter temperature - Ice Cover
Snow
Drought - Low River Level Assessment
Forest fires
Extreme winds and tornadoes
External flooding (addressed in a dedicated chapter in the ST report)
For cases in which the extreme weather conditions could affect the availability of
the off-site power supply and / or the transfer of heat to the ultimate heat sink,
based on the review of severe weather conditions and their impact on the plant,
it was concluded that these would not generate worst accident scenarios as
compared with SBO (Station Black-Out), LOUHS (Loss of Ultimate Heat Sink)
and SBO + LOUHS events.
Although none of the external events related to severe weather has the potential
to induce accident sequences not covered by the existing safety analysis, plant
operating documentation or response capacity of the Cernavoda NPP, the
specific procedure for responding to extreme weather conditions has been
revised to include more proactive actions.
Conclusions and improvement measures
TOPIC 2: Loss of safety functions
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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In the “Loss of safety functions” topic the design features have been reassessed
with regard to the fulfilment of the fundamental safety functions in a NPP:
•Control of the reactivity;
•Nuclear fuel cooling;
•Confinement of radioactivity;
•Monitoring of the critical plant parameters.
According to the “stress test” specifications, the assessment results had to
describe the means for maintaining fundamental safety functions in case of:
• loss of power,
• loss of ultimate heat sinks,
and the combination of these scenarios, following external initiating events
considered (earthquake, flooding, extreme weather conditions).
The licensees had to identify the design provisions, the unconventional (mobile)
means, the alternative fuel and water resources and the operating procedures
that can be used in order to ensure the fulfilment of the fundamental safety
functions.
Loss of Safety Functions
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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For the “Loss of Power” events, the following scenarios have been analyzed: • Loss of off-site power sources
• Loss of plant generator
• Loss of ordinary back-up generators (diesel generator, gas turbine, etc.)
• Loss of the other diverse back-up sources.
Sequential loss of these sources has been considered in the assessments
performed by the licensees.
For the “Loss of Ultimate Heat Sink” events, the following scenarios have been
analyzed: • Loss of primary Ultimate Heat Sink (UHS) (i.e. access to the water from river or sea,
that represents the primary UHS)
• Loss of primary UHS and the alternate UHS.
In addition, the combination of these events has been considered in the
assessment performed by the licensee, including as a consequence of extreme
external hazards (earthquake or flooding).
Loss of Safety Functions (cont’d 1)
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Cernavoda NPP Electrical power supply system description
There are five levels of electrical power supply at the Cernavoda NPP:
1. Class IV electrical power supplied from the grid or from the plant turbine
generator;
2. Class III electrical power supplied from the first set of diesel generators (SDG)
with 100% redundancy built-in;
3. Class I / II electrical power supplied from batteries for 8 hours;
4. Emergency Electrical Power Supplied (EPS) from the second set of diesel
generators (seismically qualified) known as emergency power supply (EPS)
designed to 100% redundancy and separation requirements;
5. The mobile diesel generators (MDG)
- Two MDG of 1.2 MWe for both Cernavoda NPP units
- Smaller MDG for other purposes (ensure the power for the fire pumps)
Loss of Electrical Power Supply
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Taking into account the design basis, the operating procedures and accident
management measures, the plant units have a high level of defense against the
loss of power and its consequences.
The cliff edge effect for SBO scenario has been defined as the plant
parameter of loss of inventory to all SGs as a heat sink. Design and operational
features, which can avoid these cliff edge effects, are available at the plant:
Eight main steam safety valves (MSSVs) can be open, by the operator or on
the auto-depressurization logic long before the dry-out condition, allowing
timely make-up of water to the steam generators. The MSSVs must remain
blocked open for long term cooling.
After the SGs are depressurized, gravity-fed make-up from the dousing tank
is available through the Boiler Make-up Water system (BMW) pneumatic
valves (PVs) in series, in the medium term.
If the PVs are uncontrolled, the time available for the make-up inventory for
decay heat removal from PHT is 27 hours. If the PVs are controlled, more
than 7 days are available for decay heat removal from PHT.
In the long term, MDGs can power the EWS pumps supplying water to the
SG (or EPS).
Conclusion on the adequacy of protection against
loss of electrical power (cont’d 1)
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Cernavoda Ultimate Heat Sinks (UHS):
The primary Ultimate Heat Sink (UHS) is based on decay power heat removal
using forced cooldown circulation in PHT system.
• PHTS (Primary Heat Transport System)+ SDCS (shutdown cooling system
) + RCW (Recirculating Cooling Water ) + RSW (Raw Service Water) +
Danube River.
The alternate heat sink is based on decay power heat removal using natural
circulation in PHT system.
• PHTS (Primary Heat Transport System)+SGs (Steam Generators) +
“feedwater train” inventory + MSSVs (Main Steam Safety Valves) +
Atmosphere.
The alternate Ultimate Heat Sink is based on decay power heat removal using
natural circulation in PHT system.
• PHTS (Primary Heat Transport System)+SGs (Steam Generators) + BMW
(Boiler Makeup Water – from Dousing Tank)/EWS (Emergency Water
Supply – from Danube River) + MSSVs (Main Steam Safety Valves) +
Atmosphere.
Loss of the decay heat removal capability/Ultimate Heat Sink
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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The assessment performed by the licensee demonstrated that there are
sufficient means (especially the dousing tank inventory) to ensure the fulfillment
of safety functions even in the case of total loss of UHS.
The cliff edge effect for this scenario is defined as the plant parameter of loss of
inventory to all SGs as a heat sink. There are sufficient design provisions to
avoid these effects.
SGs inventory will last for 2 hours prior to SGs reaching the dry-out condition.
8 MSSVs will be open based on the auto-depressurization logic long before the dry-out
condition allowing timely make-up of water to the steam generators. The MSSVs must
remain block open for long term cooling.
After the SGs are depressurized, gravity-fed make-up from the dousing tank is
available through two sets of EWS pneumatic valves in series, in the medium term.
If the PVs are uncontrolled, the time available for the make-up inventory for decay heat
removal from PHT is 23 hours.
If PVs are controlled, 7 days continuous heat sink is available.
In the long term, means to make-up water supply to the dousing tank (via on-site fire
water trucks, firewater, water from the river, etc.) either through the EWS or the ECC
would ensure long term cooling.
Conclusions on the loss of the primary ultimate heat sinks(UHS)
(primary and alternate UHS)
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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The station response at the loss of primary UHS event coincident with
– or generated by – a SBO, is based on the EWS system, being
powered from the mobile DGs.
For the unavailability of the EWS system (loss of Alternate UHS), the
response consists of fire water trucks connected directly to EWS pipes
through special connections.
The mobile diesel generators can be deployed within 2.5 to 3 hours to
restore the power to EWS.
Loss of the primary ultimate heat sink,
combined with station black out
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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Cernavoda NPP performed the assessment of the scenarios required
by “stress test” specifications, including the assessment of the cliff
edge effects;
With respect to electrical supplies, there is a good level of redundancy
and diversity.
The primary and alternative heat sinks provide a good level of
redundancy and diversity.
Following the topical peer review and the country visit, the option of
recharging the batteries or the installation of a supplementary
uninterruptible power supply for the SCA is being considered by the
licensee as a potential improvement.
Conclusions and improvement measures
TOPIC 3: Severe Accident Management
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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The assessment of the severe accident management area addressed
the following aspects:
- Current accident management measures in place for preventing and
mitigating core damage;
- Accident management provisions and design features for protecting the
containment after core damage;
- Prevention of H2 deflagration and detonation, prevention of containment
overpressure (in slow overpressurization scenarios), prevention of re-
criticality, prevention of high pressure core melt ejection scenarios,
prevention on basemat melt-through;
- Accident management for scenarios involving loss of cooling to the spent
fuel pool;
Reassessment of Severe Accident Management
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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The SAMGs for Cernavoda NPP have been developed based on the
generic CANDU Owners Group (COG) SAMGs for a CANDU-6 type
of plant. In developing the generic SAMGs, COG adopted the
Westinghouse Owners Group (WOG) approach, with the necessary
technical modifications suitable for implementation in CANDU plants,
based on extensive CANDU specific severe accident analysis and
research.
Preparation of plant-specific SAMGs was done by customisation of
the generic COG documentation package for Cernavoda NPP,
removing extraneous information not applicable to the station,
incorporating station-specific details and information and making any
other adjustments required to address unique aspects of the plant
design and/or operation.
Accident management measures in place for preventing and
mitigating core damage
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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EOPs & SAMGs
APOP Scope of application
APOP-00 SS / CRO Transient Response Strategy
APOP-E01 Dual Computer Failure
APOP-E02 Loss Of Feedwater
APOP-E03 Loss Of Instrument Air
APOP-E04 Loss Of Service Water
APOP-E05 Loss Of Class IV Power (i.e. loss of off-site power)
APOP-E06 Large LOCA
APOP-E07 Small LOCA
APOP-E08 Steam Generator Tubes Failure
APOP-E09 Partial Loss Of Class IV Power
APOP-E10 Danube Very Low Level
APOP-G01 Generic Heat Sink (MCR – Main Control Room)
APOP-G02 SCA (Secondary Control Area) Operation
APOP-G03 Station Black Out
APOP-G04 Abnormal Spent Fuel Bays Cooling Conditions
SAMG Priority Scope of application
Severe Accident Guidelines
(SAG)
SAG-1 Inject into Heat Transport System
SAG-2 Control Moderator Conditions
SAG-3 Control Calandria Vault Conditions
SAG-4 Reduce Fission Product Release
SAG-5 Control Containment Conditions
SAG-6 Reduce Containment Hydrogen
SAG-7 Inject into Containment
Severe Challenge Guideline
(SCG)
SCG-1 Mitigate Fission Product Release
SCG-2 Reduce Containment Pressure
SCG-3 Control Containment Atmosphere Flammability
SCG-4 Control Containment Vacuum
Accident management measures in place for preventing and
mitigating core damage
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
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EOPs & SAMGs
Accident management measures in place for preventing and
mitigating core damage
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Conclusion and improvements
in the area of Severe Accident Management
After the Fukushima Daiichi accident, a complex safety review of
the design was undertaken aimed to assess robustness and to
identify solutions to increase the protection against severe
accidents;
Several design improvements have been identified and have been
implemented or are under implementation to maintain fuel cooling
during severe accident conditions and to enhance the capability to
maintain containment integrity in case of severe accidents.
The status of the implementation of the improvements is presented
on the following slides.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Design improvements
in the area of Severe Accident Management - implemented
Actions implemented:
Installation of PARs (passive autocatalytic recombiners) for hydrogen
management in Cernavoda Units 1 and 2.
Accident management provisions for events in the spent fuel pools
(natural ventilation for vapours and steam evacuation, seismically
qualified fire-water pipe for water make-up).
Improvement of the reliability of the on-site emergency control centre
(increased seismic robustness) and of the on-site emergency
organization.
Installation of Special Communication Service phones in each Main
Control Room and Secondary Control Areas.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Design improvements
in the area of Severe Accident Management – in progress (1)
Actions in progress:
Additional instrumentation for severe accident (SA) management
(e.g. hydrogen concentration monitoring in different areas of the
reactor building): to be completed in 2013.
Installation of dedicated emergency containment filtered venting
system for each NPP unit: to be completed in 2013 for Unit 1 and in
2014 for Unit 2.
Improvement of the existing provisions to facilitate operator actions
to prevent a severe accident in the spent fuel pool (water level and
temperature monitoring from outside the SFP building): to be
completed by the end of 2014.
Improvements to the reliability of existing instrumentation by
qualification to severe accident conditions and extension of the
measurement domain: to be completed in 2014 for Unit 1 and in
2015 for Unit 2.
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Design improvements
in the area of Severe Accident Management – in progress (2)
Actions in progress:
Implementation of a design modification for water make-up to the
calandria vessel and the calandria vault (completed for Unit 2
calandria vessel): to be finalized in 2013;
Establishment of a new seismically qualified location for the on-site
emergency control centre and for the fire fighters; this location will
include important intervention equipment (mobile DGs, mobile diesel
engine pumps, fire-fighter engines, radiological emergency vehicles,
heavy equipment to unblock roads, etc) and will be protected against
all external hazards: to be completed in 2015;
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
44
Accident management for scenarios involving
loss of cooling to the SFB
In response to the Fukushima accident, based on WANO SOER 2011-2
recommendations, an emergency operating procedure called “APOP G04 -
Spent Fuel Bay cooling abnormal conditions“ was developed, validated and
issued in order to address prolonged/ extended loss of Spent Fuel Bay cooling
capability, main goal being to prevent fuel bundles damage and H2 generation,
due to overheating.
During the worst case scenario (loss of Class IV together with loss Class III,
earthquake or Station Black-Out) when the normal cooling and normal
demineralized water make-up is lost for a prolonged period of time (there are
60h available until water in the bays would start to boil), the APOP G04
procedure guides the operators to establish another means of water make-up in
the Spent Fuel Bay. This is performed using fire trucks or mobile pump via hose
connections, maintaining spent fuel submerged.
To support APOP G04 execution, several design changes and operational
measures have been implemented. It should be noted that on the assumption
than no action is taken to replete the SFB inventory, there would be 15 days
until the first row of fuel bundles become uncovered.
Conclusions
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
46
Design measures: a set of plant modifications have been proposed by the
Cernavoda NPP to be implemented in order to:
Increase the robustness of the plant and avoid the occurrences of severe
accidents following extreme external hazards
Provide different means for the mitigation of severe accidents effects, to
ensure the integrity of the containment and limit the radioactive releases in
case of beyond design basis accidents
To ensure the monitoring of the critical plant parameters and the emergency
response in case of severe accidents determined by extreme external
hazards
Operational measures: review of the abnormal and emergency procedures as
well as Severe Accident Management Guidelines in order to improve the
response of the operators and technical personnel of the plant in case of
extreme external conditions and consequential loss of power/loss of UHS;
training of the personnel in relation with the new operating procedures and
SAMGs.
Stress Test Results (1)
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
47
Some of the proposed modifications or reviews have been already
implemented, others are in progress, approved by CNCAN.
The Romanian National Action Plan is summarised in the annex 2 of
the sixth revision of Romanian National Report under the Convention
on Nuclear Safety (http://www.cncan.ro/assets/stiri/Romanian-Report-
for-the-CNS-6th-Edition.pdf) and provides an outline of the main
improvement activities resulting from the post-Fukushima safety
reviews performed to date.
The action plan has been developed for bringing together the actions
identified from regulatory reviews, self-assessments, peer-reviews
and generic recommendations at international level.
All major design improvements are planned to be completed by the
end of 2015.
Stress Test Results (2)
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
48
As a follow-up to the “stress test”, CNCAN will continue the following
actions:
Verification of the licensee’s implementation of the action plan;
Verification of the implementation of a surveillance and testing programme
of the licensee for the equipment procured to mitigate the consequence of
SBO, loss of UHS and mitigation of the severe accidents;
Verification of the operators’ training programme to include the SAMGs;
Update of the routine inspection package of CNCAN resident site
inspectors to include the verification of the availability of the equipment
procured to mitigate the consequence of SBO and loss of UHS;
Requirements on severe accident management will be included in a new
regulation that is under preparation for publication;
Verification of the implementation of the design modifications which are in
progress and of the update of the relevant operating documentation.
Further Regulatory Actions
Examples of safety improvements
after Fukushima
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Flooding resistant doors
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Main Steam Safety Valve manual opening kit
hydraulic jack
N2 supplies for
MSSVs pneumatic
actuation
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
Mobile Diesel Generators
Technical Meeting on Evaluation of Nuclear power Plant Design Safety in the Aftermath of the Fukushima Daiichi Accident, Vienna, Austria, from 26 to 29 August 2013
53
Thank you for your attention