Transcript (STUK). - Varans.vn

STUK- independent Regulatory Body for
nuclear and radiation safety in Finland and
the use and role of TSO’s
VN/RA/01 Task 1&2 Workshop
Hanoi, October 2012
Confidential
Ilari Aro
Ref. to Ilona Lindholm, Keijo Valtonen
STUK
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Nuclear power programme in Finland
Fennovoima Ltd
• New utility, no operating reactors
• DiP approved for FA1, Hanhikivi Site
Loviisa NPP (Fortum)
Olkiluoto NPP (TVO)
• 2 operating units - ABB BWRs
• OL3 (EPR) under construction
• DiP approved for OL4
• Interim Spent Fuel Storage at site
•L/ILLW repository
• Posiva “Onkalo”
• 2 operating units – VVERs
• Interim Spent Fuel Storage at site
• L/ILLW repository
Photo: Fortum
Photo: TVO
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK - Radiation and Nuclear Safety Authority
Mission:
Protecting people, society,
environment, and future
generations from harmful
effects of radiation
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Organisation
• Regulatory Body
– use of nuclear reactors
– nuclear waste and materials
– use of radiation
• Research Centre
– health effects of ionising and nonionising radiation
– natural radiation - occurrence and
prevention
– environmental research
– radiation threats and
preparedness for accidents
– dosimetry and metrology
– medical use of radiation
– non-ionising radiation
Figures indicate staff number (356) at the end of 2011.
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
• Expert organisation
– national preparedness for
radiation accidents
– information and advice to public
and other authorities
– contracted expert services
– radiation measuring and
calibration services
Organisation of Nuclear Reactor Regulation
Nuclear Reactor Regulation
Director Petteri Tiippana
Deputy director
Marja-Leena Järvinen
Oversight support
Nuclear Security
Department services
Nuclear facilities and
systems
Assistant director
Keijo Valtonen
Structures and components
Assistant director
Martti Vilpas
Projects and operational
safety
Assistant director
Tapani Virolainen
Reactor and safety
systems
Section head
Risto Sairanen
Mechanical engineering
Section head
Petri Vuorio
Organisations and
operation
Section head
Jukka Kupila
Risk analysis
Section head
Reino Virolainen
Civil engineering
Section head
Pekka Välikangas
Projects
Section head
Kirsi Alm-Lytz
Electrical and
automation systems
Section head
Kim Wahlström
Manufacturing
technology
Section head
Juhani Hinttala
Radiation protection
Section head
Olli Vilkamo
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
5
Project managers:
OL 3: Tapani Virolainen
New projects: Janne Nevalainen
LARA: Kaj Söderholm
Main organizations involved in licensing and
safety assessment in Finland
Government
Ministry
Licenses
STUK
Safety assessment
Statement on nuclear safety
Main TSO: VTT
Safety analysis
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
6
Typical TSO’s in Finland for safety analysis
and research
• VTT – Technical Research Centre of Finland
– Branches for all technical fields of society
– For nuclear and waste safety about 100 experts (altogether
about 2000 staff)
– Main TSO for STUK
• Geological Institute and seismological institute for siting
studies
• Meteorological institute for meteorology and emergency
response
• Universities (e.g. Helsinki, Aalto and Lappeenranta
universities) e.g. radiochemistry, nuclear waste,
thermohydraulics and severe accident research
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
23 August 2010
7
Review and Assessment by the Regulatory Body
– Licensee is responsible to demonstrate safety and fulfilment of safety
requirements (requirements are presented in the YVL guides)
– STUK conducts independent review and assessment in licensing steps
(before license is granted) and before modifications are implemented
at the plants
– Focus and scope of STUK’s review and assessment depends on the
licensing phase
– Safety assessment tools such as deterministic and probabilistic safety
analyses are utilised in STUK’s review and assessment
– STUK has established internal guidance for review work to ensure
consistent review and assessment process and application of graded
approach
– Technical Support Organisations are utilised in specific areas (e.g.
comparative accident and transient analyses), but STUK makes
decisions on the safety case
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
DiD-levels, event categories and frequency of
events belonging to each category
Level 1
Normal operation (DBC 1)
Level 2
Anticipated operational
occurrences (DBC 2)
Level 3a
Class 1 postulated accidents 10-2/a > f > 10-3/a
(DBC 3)
f > 10-2/a
Class 2 postulated accidents f < 10-3/a
(DBC 4)
Level 3b
Design extension conditions
(DEC)
CCF
rare events
Level 4
Severe accidents (SA)
safety goals
CDF <10-5/a, LRF < 5x10-7/a
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Ilari Aro, 31 October 2010
9
Acceptance criteria for radioactive
releases / max doses to general public
• DBC 1, Normal operation
– radiation dose limit 0,1 mSv / year for the entire site
• DBC 2, Anticipated operational occurrences
– radiation dose limit 0,1 mSv
• DBC 3, Class 1 postulated accidents
– radiation dose limit 1 mSv
• DBC 4, Class 2 postulated accidents
– radiation dose limit 5 mSv
• DEC, Design extension conditions
– radiation dose limit 20 mSv
• SA, Severe accidents
– release < 100 TBq Cs-137 equivalent
– no acute health effects
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Ilari Aro, 31 October 2010
10
Acceptance criteria for fuel
•
•
DBC 1 , Normal operation
DBC 2 , Anticipated events
•
•
DBC 3 “Class 1” postulated accidents
•
•
number of rods in heat transfer crisis < 1%, PCT < 650 °C, and extremely low
probability of fuel damage by the mechanical interaction between fuel and cladding
DBC 4 ”Class 2" postulated accidents
•
•
95/95 confidence with respect DNB or dry-out, no (internal) fuel melting, nor damage
due to pellet-cladding mechanical interaction.
the higher the frequency of a postulated accident, the smaller the number of
damaged fuel rods. Number of damaged fuel rods < 10%. Max PCT < 1200 C.
Limited enbritlement. Enthalpy limit 140 cal/g for failure (230 cal/g not be exceeded).
Enthalpy limits are valid for fuel burnups up to 40 MWd/kgU. Limits for higher burnups
shall be justified by experiments. No danger to long-term coolability
DEC, Design extension conditions
•
•
DEC A
– Max PCT < 1200 C. Limited enbritlement. Enthalpy limit 140 cal/g for failure
(230 cal/g not be exceeded). No danger to long-term coolability
DEC B,
– Max PCT < 1200 C. Limited enbritlement. Enthalpy limit 140 cal/g for failure
(230 cal/g not be exceeded). No danger to long-term coolability
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK
September 2011/ I Aro
11
10 % FUEL FAILURE LIMIT
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK
September 2011/ I Aro
12
Analyses of plant behaviour: Examples of
cases to be analysed
•
Examples of initiating events to be analysed are:
DBC 2
– disturbance in the reactor power control or other disturbance, which
causes a change in reactivity
– disturbance in primary circuit flow, pressure control or water volume
control
– disturbance in steam pressure or steam flow
– disturbance in feedwater flow or feedwater temperature
DBC 3,4
– leaks from the primary circuit during power operation, change in
operational state, refuelling and/or outage
– leak from secondary circuit (PWR)
– leak from primary to secondary circuit (PWR)
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK
September 2011/ I Aro
13
Design Extension Conditions (DEC)
•
DEC A
– includes conditions in which a common cause failure (CCF) in a safety system is
assumed during anticipated operational occurrence (DBC 2) or class 1 accident
(DBC 3), overall frequency of an event ~10-4 - 10-5
- as an example
-
ATWS
station black out
total loss of feed water
LOCA together with the complete loss of one emergency core cooling system
total loss of the CCWS
total loss of the RHR
loss of ultimate heat sink
loss of fuel pool cooling
– realistic assumptions are applied for accident analysis
•
•
single failure is assume in safety systems
DEC B
– includes complex sequences and rare external events
– as an example
•
•
•
multiple stem generator tube rupture (~10)
extreme weather condition
large airplane crash
– realistic assumptions are applied for accident analysis
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK
September 2011/ I Aro
14
Analyses of plant behaviour: Examples
of severe accidents
•
Severe accident analyses shall be used to study factors which affect
containment integrity, leak tightness and the operability of
containment systems. They could include i.e.:
– total, long lasting loss of AC power
– total loss of feedwater
– leak of primary coolant without emergency cooling during power
operation or a maintenance, refuelling or other outage
– leak of primary coolant and blockage of coolant recirculation
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK
September 2011/ I Aro
15
Analyses of releases and radiation
doses : Examples of postulated
accidents
•
Separate radiation dose analyses shall be made if the dose upper
limit cannot be concluded from the results of other analyses. Some
examples are:
– Large leak of coolant from the primary circuit during power operation. A
typical example of accidents during which radioactive substances are first
released into the containment and gradually leak out.
– Leak of reactor coolant outside the containment due to an instrument line
rupture
– Leak from steam generator primary to secondary side. The total rupture
of one or multiple steam generator tubes shall be analyzed by assuming
that also the safety valve of the steam generator has stuck open in a
case it is expected to open. Also a leak larger than the one mentioned
above shall be analyzed if estimated possible on the basis of the
structure of the steam generator.
– Leak out of the primary circuit during a maintenance, refuelling, or other
outage.
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK
September 2011/ I Aro
16
Analyses of releases and radiation doses :
Examples of postulated accidents (continued)
– Leak outside the containment in an unisolated steam line connecting to a
steam generator in which, before the initiation of the accident, the largest
primary to secondary circuit leak (PWR) allowable in the Technical
Specifications has occurred.
– Leak in a steam line outside the containment or in a reactor coolant
purification line (BWR).
– Damage outside the containment in a system containing radioactive
gases.
– Damage outside the containment in a system containing radioactive
liquids.
– Damage of a fuel assembly which has been removed from the reactor.
– Dropping of a transfer or transport cask containing spent fuel during
hoisting, in a situation where the cask is not tightly closed, or dropping of
the fuel cask during transfer.
– Dropping of a heavy object on top of stored fuel or an open reactor.
•
Severe accidents
– Analyses shall be carried out for cases which on the basis of
containment behaviour and conditions and the concentration of
radioactive substances in the containment are estimated to cause the
most extensive releases.
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
STUK
September 2011/ I Aro
17
Reliability of the analysis methods
•
•
•
•
•
Methods of analysis mean i.a. methods based on hand calculations, computer
programs and the application of experimental data. The reliability of the analysis
methods used shall be justified. A description of the analysis methods used shall
be given, including their general principles as well as the physical models and
numerical methods used.
The experimental correlations used in the calculations shall be justified by
presenting the measurement data from which the correlations have been derived.
If the correlation is commonly known and the measurement data are publicly
available, a bibliographic reference is sufficient.
The analysis methods shall be adequately verified for the treatment of the events
in question. Both numerical methods and physical models shall be verified.
Numerical methods shall be verified by adequate reference calculations. Physical
models shall be verified by demonstrating their ability to depict suitable separate
effects tests or integral tests for complete systems or nuclear power plant
transients. In addition, comparison with other, earlier verified models may be
utilised.
If sufficiently reliable calculation methods are not available, the analysis shall be
justified by experiments. This requirement applies especially to most phenomena
essentially relating to severe accident management, for example, the long term
coolability of reactor core debris after a severe accident.
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
18
Performing safety analysis
•
•
•
•
Limiting transients or accident cases are selected from the point of view
of fuel design criteria or pressure vessel design criteria or for
containment analysis purposes. The purpose is to check that design
values are according to the regulatory requirements, input data has
been properly selected and that there are enough conservatism and
design margin in design. Also the analysis models are checked by using
comparison analyses with different analysis tools.
The ideal case is to use different calculation codes and models. In most
cases this also takes place. In this respect, STUK has been quite
successful.
VTT in Finland has the role of developing analysis tools or transform
them into the Finnish conditions, test their validity and verification and
also to perform safety analysis when needed. VTT validates the codes
e.g. by performing some benchmark calculations against experiments
or as international co-operation.
In Finland, originally, many computer codes used were of US design
and it was necessary to apply them into the reactor systems that were
different from the original purposes. For this reason, VTT was needed
for this basic design and verification work. Currently, VTT has developed
its own analysis tool system that is efficient in the safety analysis.
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
19
Performing safety analysis
• Analysis tools are mainly provided by VTT or via VTT to
STUK. STUK has also developed some computer codes by
itself: e.g. for PSA and for radiation safety calculations.
• If necessary, STUK requires from the licensee also the
analysis tools and the input data used for its review in addition
to the analysis reports and results. For the safety assessment
purposes, different computer codes and models are mainly
used.
• Power companies typically use codes developed by the plant
vendor e.g. ABB computer codes for BWR or they have
supported the development of own codes as Fortum/VTT
codes for VVER. STUK mainly uses codes developed by VTT
or which have been received through international cooperation e.g. with US NRC.
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
20
Codes used for Olkiluoto BWR
ANALYSIS
LICENSEE
REGULATOR
Core calculation
Polca / Updat
not performed
Transient analys.
Bison/Ramona(3D)
TRAB / Ramona4
LOCA analysis
Coblin / Dragon
RELAP5
Containment
Copta
CONTEMPT
Structural analys.
commercial codes
not performed
Severe accident
MAAP / MELCOR
MELCOR
Radioact. release
Meteorology
TUULET
ARANO, VALTO
PSA
SPSA
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
SILAM (FMI)
October 2012 / IA
SPSA
21
Codes used for LOVISA - VVER
ANALYSIS
LICENSEE
REGULATOR
Core calculation
HEXBU
not performed
Transient analys.
HEXTRAN (3D)
TRAB (3D)
LOCA analysis
APROS
RELAP5 / (APROS) / FRAPTRAN
Containment
APROS
CONTEMPT
Structural analys.
commercial codes
not performed
Severe accident
MAAP / MELCOR
MELCOR
Radioact. release
Meteorology
TUULETV2001/MERI2002 /
TUULET (acc)
ARANO, VALTO
PSA
RiskSpectrum / SPSA
SPSA
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
22
SILAM (FMI)
What to assess - plant level
Nuclear safety analysis
•
Plant functional design (anticipated transients, design basis
accidents, design extention conditions, severe accidents, PSA,
radioactive releases, environmental effects)
• Site evaluation and environmental aspects
Plant design - system analysis
• system design features and description - compliance with design
criteria
• safety classification, QA/QC requirements, regulatory control
• system failure analysis - redundancy, diversity, separation,
interaction with other systems, risk assessment
• system operation - functional requirements; operation during
different plant states including accidents
• environmental conditions & classification
• operational limits and conditions, surveillance and testing
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
23
Criteria for safety assessment (plant level)
• General design criteria
– basis for safety assessment review report
• YVL guides
– YVL 1.0 Safety criteria for design of nuclear power plants
– YVL 2.2 Transient and accident analyses for justification of
technical solutions at nuclear power plants
– YVL 2.8 Probabilistic safety analyses (PSA)
– YVL 6.2 Fuel design limits and general design criteria
– YVL 7.1 Limitation of public exposure in the environment of and
limitation of radioactive releases from nuclear power plants
• YVL guide 2.0 gives criteria for the design of safety systems
–
–
–
–
YVL 2.1 cover safety classification
YVL 2.7 cover failure criteria
YVL 1.4 cover QM
YVL 2.5 cover pre-operational and start-up testing of NPP
• YVL guide system provide detailed criteria for structures,
systems and components
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
24
System design requirements
– Safety classification based on PSA insights
– N+2 failure criterion for systems that deal with design basis
events; besides redundancy, also adequate diversity and
segregation
– Proven technology
• properly evaluated operational experience
• experimental demonstration & analysis (novelties, such as
“passive” systems)
– Adequate demonstration of performance and safety margins
on the basis of deterministic studies and PSA
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
25
Design basis for containment
1) Large break LOCA
• adequate capacity to carry pressure loads and to limit radioactive
releases must be shown in conditions expected after a LB LOCA
containment
• this gives a sound basis to manage also severe accidents
2) Severe accidents
• all foreseeable loads threatening the containment integrity in
connection with a severe core damage must be identified, and
necessary protection (prevention or mitigation) must be provided
against each load
– in Olkiluoto 3, pressure caused by hydrogen burn is the
limiting design event for containment
3) External events
• potential external events must be identified and protected against
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
26
Features required to mitigate severe
accidents
• Severe accident management strategy is mandated in containment
design
– high pressure failure of reactor vessel prevented by dedicated
depressurization system
– hydrogen management with autocatalytic recombiners to prevent
detonation
– low pressure melt arrested in a core catcher, with passive longterm cooling
– containment integrity against dynamic loads
– containment pressure management in long term
– containment leak tightness criteria from release limits
• AC power supply systems and I&C systems dedicated to support
severe accident management are required
• For systems dedicated for protection against severe accidents,
single failure criterion applies
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
27
Protection against external threats
After September 11, 2001: political and public will was expressed to
improve protection against terrorist actions
– Reconsideration of aircraft crash design basis
- consider large passenger and military aircrafts
- no immediate release of significant amount of radioactive
substances
- initiation and maintenance of key safety functions in spite
of the direct consequences of the event (penetration of
structures by impacting parts, vibration, explosion, fire)
– Microwave and biologic weapon consideration
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
October 2012 / IA
28
Structure of the new YVL Guides
A Safety management of a
nuclear facility
B Plant and system design
A.1 Regulatory control of the
safe use of nuclear energy
B.1 Design of the safety
systems of a nuclear facility
C Radiation safety of a
nuclear facility and
environment
D Nuclear materials
and waste
C.1 Structural radiation
D.1 Regulatory control
safety of a nuclear facility of nuclear nonproliferation
A.2 Siting of a nuclear facility
B.2 Classification of systems, C.2 Radiation protection
structures and equipment of
and dose control of the
D.2 Transport of
A.3 Management systems of a
a nuclear facility
personnel of a nuclear
nuclear materials and
nuclear facility
facility
waste
B.3 Safety assessment a
A.4 Organisation and
NPP
C.3 Control and
D.3 Handling of spent
personnel of a nuclear facility
measuring of radioactive nuclear fuel
B.4 Nuclear fuel and reactor
releases to the
A.5 Construction of a NPP
D.4 Handling of lowB.5 Reactor coolant circuit of environmental of a
and intermediate-level
A.6 Operation and accident
nuclear facility
a NPP
waste and
management of a NPP
C.4 Radiological control
decommissioning of a
B.6 Containment of a NPP
A.7 Risk management of a
of the environment of a
nuclear facility
NPP
B.7 Preparing for the internal nuclear facility
D.5 Final disposal of
and external threats to a
A.8 Ageing management of a
C.5 Emergency
nuclear waste
nuclear facility
nuclear facility
preparedness
B.8 Fire protection of a
arrangements of a NPP
A.9 Reporting on the operation
nuclear facility
of a nuclear facility
E.1 Manufacture and use of
nuclear fuel
E.2 Construction plan of the
mechanical components and
structures of a nuclear facility
E.3 Regulatory control of the
mechanical components and
structures of a nuclear facility
E.4 Verification of strength of
pressure equipment of a
nuclear facility
E.5 In-service inspections of
the mechanical components
and structures of a nuclear
facility
E.6 Buildings and structures
of a nuclear facility
E.7 Electrical and I&C
equipment of a nuclear
facility
A.10 Operating experience
feedback of a nuclear facility
A.11 Security arrangements of
a nuclear facility
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
E Structures and
equipment of a nuclear
facility
29
Computer programs for safety analysis in VTT in Finland
ENDF/B, JEF
NJOY
CASMO libraries
CASMO 4
SIMULATE
rectangular
Nuclear Data
Assembly wise
group constants
TRAB-3D
BWR
Steady state
and transient
fuel behaviour
RELAP5/
Mod3
Thermal hydraulics
Plant analyzer
Training simulator
HEXBU-3D
hexagonal
Reactivities, power and burnup
ENIGMA, FRAPCON,
SCANAIR, FRAPTRAN
FRAPTRAN-GENFLO
APROS
simulation
environment
Thermal
hydraulics
Severe accidents
MELCOR
CONTAIN
SCDAP/RELAP5
Integral severe
accident analyses
Containment
performance
Core and primary
system
HEXTRAN
VVER
Transient and accident analyses with
3D core model, complete circuit models
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
PASULA
Pressure vessel
integrity in severe
accidents
Reactor analysis calculation system
B a sic n u cle a r d a ta
E N D F /B , J E F
= c od es d e velo pe d b y V T T
= c od es partly d evelo p ed by V TT
N u cle a r d a ta p ro ce ssin g
= c od es a p plied b y V TT
NJO Y
N u cle a r d a ta lib ra rie s
(2 5 - 7 0 e n e rg y g ro u ps )
C A S M O lib ra rie s
Ste a dy sta te fu el
rod b e h aviou r (a lso
p ro b a b ilistic a n alyses)
ARES
E N IG M A ,
FRAPCON
C A S M O -4
C A S M O -H E X
S q u are h exag o n al
h ex ag o n al
S IM U L AT E
H E X B U -3 D
sq u are
V VE R
APR OS
PWR
BWR
V VE R
S C A N A IR ,
F R AT R A N
F R A P T R A NG EN FLO
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
TRAB
S M AT R A
BWR
PWR
T R A B --3D /
SM ABRE
BWR, PWR
PVM/NN
C a lcu la tio n o f re a ctivitie s,
p o w e r a n d b u rn u p
d istrib u tio n s e tc .
D a ta tra n sfe r a n d
co n d e n sa tio n fo r o n e d im e n sio n a l g ro u p co n sta n ts
CROCO
F ue l ro d b e h a vio ur
d u rin g R IA s
and LO CAs
C a lcu la tio n o f
a sse m b lyw ise tw o
g ro u p co n sta n ts
H EX TR A N /
SM ABRE
V VE R
31
O n e -d im e n sio n a l d yn a m ic s
co d e s
T h re e -d im e n sio n a l
d yn a m ics co d e s
Example on coupling of safety codes used in accident analysis
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
PVM/NN
32
Code development in VTT
V T T h a s a co m p re h e n sive n u cle a r re a cto r
sa fe ty a n a lysis co m p u te r co d e syste m . It
p ro vid e s a co m p le te se t o f to o ls fro m
b a sic n u cle a r d a ta to th re e -d im e n sio n a l
tra n sie n t a n a lyse s a n d se ve re a ccid e n ts.
V T T E n e rg y h a s a lo n g e xp e rie n ce o f
sa fe ty a n a lyse s fo r V V E R a n d B W R re a cto rs in clu d in g a ll typ e s o f p o stu la te d
a ccid e n ts. T h e ke y co d e s, T R A B -3 D fo r
B W R s a n d P W R s, H E X B U -3 D a n d
H E X T R A N fo r V V E R s a n d A P R O S
m u ltip u rp o se to o l, h a ve b e e n d e ve lo p e d
a t V T T , th e A P R O S co d e jo in tly w ith
F o rtu m . R e ce n tly a lso th e U S N R C fu e l
tra n sie n t co d e F R A P T R A N h a s b e e n
a p p e n d e d w ith th e g e n e ra l p u rp o se flo w
m o d e l G E N F L O , d e ve lo p e d b y V T T .
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
PSA tool in use in STUK
•
FinPSA is a comprehensive risk and reliability analysis
tool intended for full scope PSA/PRA modeling. The tool
has been developed and maintained since 1988 by
Radiation and Nuclear Safety Authority of Finland
(STUK). The tool is designed to support the main
activities related to PSA/PRA by easy model creation,
efficient and versatile analysis, good traceability, flexible
reporting and information exchange capabilities. All
these features make your work comfortable in living PSA,
plant assessment and operational modifications. From
the beginning of 2012, STUK and VTT (Technical
Research Centre of Finland) have initiated a project to
develop the tool ahead and to train new FinPSA experts.
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
23 August 2010
34
TSO Support for Olkiluoto 3 Review
Criteria for TSO’s;
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–
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Competent organisation to carry out EPR analysis
Independent from the licensee
Codes independent from the licensee’s codes
Adequately validated codes for EPR
Main organisations used as TSO’s were
– Technical Research Centre of Finland (VTT)
– Institute for Safety and Reliability (ISar)
Main computer codes used by TSO’s
– TRAB3D/SMABRE (VTT)
– APROS (VTT)
– MELCOR (VTT)
– ATHLET (ISar)
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
TSO Support for Olkiluoto 3 Review
Main Finnish organisations used were
Technical Research Centre of Finland (VTT) and
Lappeenranta Technical University (LTU)
VTT:
• Transient analyses with the TRAB3D/SMABRE code
• Design Basis Accident analyses with the APROS code
– Independent comparative analyses of primary circuit
behaviour during accident situations
» Small break LOCA
» Steam Generator Tube Rupture
» Large break LOCA
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Tools for Severe Accident Assessment
• Integral severe accident analysis tools
– Used for assessment of progression of whole accident scenario
– Good for assessment of overall performance of ESF
– Applicable also for PRA level 2 studies
• Detailed, separate effects severe accident codes
– State-of-the-art models
– Used for in-depth studies of safety critical SA phenomena
– Generally related to phenomena that cannot be modelled/solved
with the methodology used in integral systems analysis codes
• Experiments
– Last resort to support or validate analytical results
– Equipment survivability studies
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Analytical tools
• Integral system codes obtained from abroad
• Detailed, separate effects SA codes obtained from abroad or
developed at VTT
• Experiments:
– Participation in major international experimental research
programmes
– Small-scale experiments/testing also in-house
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Integral Severe Accident Analysis Tools
at VTT
•
•
MELCOR (version 1.8.6 and 2.1) for integral plant analysis, source term,
developed in Sandia National Lab’s for USNRC
– Main severe accident analysis tool since 1990 at VTT
– Validation: Sandia’s validation and QA
– PCCS condenser application tested against:
– Concrete erosion model validated against various OECD/MCCI
experiments
– Plant models for Loviisa 1&2, Olkiluoto 1&2, Olkiluoto 3
ASTEC Integral plant analysis, source term, developed by IRSN
(France)/GRS (Germany)
– Participation in ASTEC users group for validation efforts since 2011
– Only RPV lower head melt pool studies performed so far
– Applications of fission product source term models under way
– Long-term goal to build full plant inputs for Finnish NPPs
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Separate Effects Severe Accident Analysis
Tools at VTT (1/4)
•
PASULA code suite for detailed mechanical analyses of structures
– Creep rupture of lower head
– Lower head penetration integrity
– Ex-vessel cooling of lower head
– Validation against several Sandia’s LHF and OLHF tests
– Applied for Loviisa 1 & 2, Olkiluoto1 & 2 and SWR-1000 reactors
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Separate effects Severe Accident
Analysis Tools at VTT (2/4)
•
•
•
•
CORQUENCH for concrete erosion, melt pool coolability, developed at
Argonne National Lab for USNRC
– Code tested at VTT against various CCI-tests and COMET-tests
FLUENT (CFD code) for hydrogen mixing and deflagration combustion,
commercial UK code
– Tested at VTT against THAI-experiments for mixing
– Tested at VTT against FLAME tests for flame speed and
acceleration
– Applied for Olkiluoto 1,2,3 and Hungarian Paks (VVER 440)
TONUS for hydrogen detonation, developed at CEA, France
– Tested at VTT against ENACEFF tests for flame acceleration
– Tested at VTT against FLAME experiments for detonation
DET3D for hydrogen detonation, developed at FZK, Germany
– Interface with ABAQUS structural analysis code
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Separate effects Severe Accident
Analysis Tools at VTT (3/4)
•
•
•
•
TEXAS-V for steam explosions, developed at UW, USA
– Validated at VTT against KROTOS experiments KS-1 and KS-2
Applied for Olkiluoto 1 & 2
MC3D V3.5 for steam explosions, developed at IRSN, France
– Validation at VTT ongoing against KROTOS and TROI tests
– Applied for Olkiluoto 1&2
MEWA for corium particle bed coolability, developed at IKE Stuttgart
– Validation at VTT against own STYX and COOLOCE experiments
– Applied for Olkiluoto 1 & 2
CONTAIN 2.0 for containment T/H and source term analyses,
developed at Sandia National Lab for USNRC
– Validation at VTT against Fortum’s VICTORIA tests, HDR tests
– Applied for Loviisa 1 & 2
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Separate effects Severe Accident
Analysis Tools at VTT (4/4)
• CHEMPOOL for calculation of pH in containment pools,
developed at VTT
– Validated against a matrix of titration experiments
• QA’ed for application in safety related work to US
– Used for supporting design calculation for design of pH control
system of OL1/OL2
– Applied also for Olkiluoto 3 and ESBWR (GEH)
• RADTRAD for dose calculations for DBA cases (NRC
Regulatory Guide 1.183); developed for USNRC
– Applied for Ringhals 2 (Westinghouse PWR)
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Application example: Olkiluoto unit 1 and 2
BWR 860 MW
•
•
•
•
•
•
•
•
RPV failure mode
Effects of recriticality
Hydrogen
Core debris coolability
Steam explosions
pH control
Elastomer survivability
Painted wall structure decontamination
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
OLKILUOTO 1 and 2 BWR 860 MW
•
•
•
•
•
•
•
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Containment inerted with nitrogen
 in-containment hydrogen
combustion risk low
Hydrogen generation 
contributes to containment
pressurization and H2 leak to RB
Recriticality
RPV is assumed to fail  debris
coolability and stabilization in the
Lower Drywell crucial
Flooding of Lower Drywell prior to
Pressure Vessel failure  particle
bed coolability is an issue
Ex-vessel steam explosions
Containment pool pH conrol
RPV lower head failure mode
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
RPV lower head penetrations
Control rod penetration
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
Instrument tube
penetration
MCP penetration
Penetration failure
•
Liquid corium,
T=2550 K
•
Weld
•
Penetration weld fails by loss
of strength in about 40 s after
contact with melt
Instrument tube inside nozzle
tube (”blue tube”) can fall
downward and open a flow
path
Uncertainties:
–
–
RPV wall
Instrument
tube
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
•
•
Falling tube becomes stucked at
lower elevation
Refreezing of melt blockage in
discharge channel
Control rod penetration would
fail 5.5 hours after melt arrival
RCP opening would fail 5.2
hours after arrival of melt
Creep rupture of lower head
•
•
•
•
•
SÄTEILYTURVAKESKUS • STRÅLSÄKERHETSCENTRALEN
RADIATION AND NUCLEAR SAFETY AUTHORITY
If corium is not able to discharge
through a failed penetration
Heat fluxes from the melt pool to
the RPV wall obtained from
MELCOR calculation
Creep rupture occurs near the
interface of metallic and oxidic
melt layer at 5.5 hours after melt
arrival
Practically no difference in timing
between RPV wall creep rupture,
control rod tube failure and RCP
opening failure
But, the initial hole areas may
differ