Side Event of IAEA General Conference

Severe Accident Analyses ofFukushima-Daiich Units 1 to 3
Harutaka Hoshi and Masashi Hirano
Japan Nuclear Energy Safety Organization (JNES)
September 17, 2012
Side Event by Government of Japan
at 56th IAEA General Conference, Sep 17, 2012
1Contents
1. Background
2. Accident Progression at Unit 1
3. Accident Progression at Unit 3
4. Hydrogen Explosions at Units 1 and 3
5. Source Terms
6. Environmental Consequences
7. Summary21

1. Background• The SA progression analyses done by TEPCO and JNES
were reported to the IAEA Ministerial Conference in June
2011.
• Since then, the analyses have been continuously improvedby taking into account new information on, such as:– Operation of isolation condensers, RCIC, venting
system, etc.
– Leakage / Failure of primary pressure boundary andcontainment, etc.
• However, it is still difficult to predict when and how much
molten core fell into containment mainly due to large
uncertainty in injection water flow rate into the core.
• Very recently, a preliminary analysis of migration anddeposition of radioactive materials in the environment byusing the radioactive release rates calculated by the SA
progression analysis.
3

2. Accident Progression at Unit 1
Major improvements:• In the previous analysis, leakage from primarypressure boundary due to over-temperature was
not considered. Therefore, the primary systemdepressurization took place at the timing oflower head failure (melt through). In the present
analysis, the leakage due to over-temperature
was considered.
• Similarly, the leakage due to over-temperature
was considered at the top-head flange of thecontainment (PCV).
42

Unit 1

RPV Pressure inPrevious Analysis (June 2011)○△ Plant recordsReactor pressure vessel (RPV)
lower head failed and RPVpressure rapidly decreased.

In this calculation, leakage due to overtemperature was not considered. Therefore,
the primary pressure boundary remained
intact although it’s temperature exceeded
720K. This was unrealistic.

Reference: Nuclear Emergency Response Headquarters Government of Japan, "Report of Japanese Government to the IAEA Ministerial
Conference on Nuclear Safety - The Accident at TEPCO's Fukushima Nuclear Power Stations - ," June 2011
5Assumption of leakage from primary pressure
boundary due to over-temperature• In the present analysis, it is presumed
that the leakage took place:– At the flange of safety relief valve (SRV)
when it’s temperature exceeded 720K,– At the in-core instrumentation tubes when
the peak cladding temperature (PCT)
exceeded 1000K.
SRV flange~720 KSRV

SRM/IRM
dry tube
TIP dry tube

SRM/IRM, TIP
PCT > 1000 K

S/C

SRM: Source Range Monitor
IRM: Intermediate Range Monitor
TIP: Traversing In-Core Probe
Reference: http://www.tepco.co.jp/en/nu/fukushima-np/images/handouts_120312_04-e.pdf
63Potential Leak locations of Mark-I Type Containment• It is presumed that a leakage took place at the top head flange ormachine hatch when it’s temperature exceeded 620K.
Unit 1

Unit 3

Causes and Countermeasures:The Accident at TEPCO’s Fukushima NPS,
Masaya Yasui, NISA/METI, March, 20127

Unit 1

RPV Pressure in Present Analysis:
Results became more realistic.
○△ Plant records

Depressurization
due to leakage
In the present analysis, RPV
lower head failed after RPV
depressurized.

RPV Lower head failed.
Leakage occurred atprimary pressureboundary.

84

Unit 1

PCV Pressure in Present Analysis:
Better agreement was obtained.

Increase in PCV pressure due
to steam and hydrogen leaked
from RPV
2nd vent also
succeeded
partially○ Plant records

PCV leakage started
due to over temperature
1st vent partiallysucceeded

9Molten Core Concrete Interaction (MCCI)• JNES performed cross-checking of the calculation done byTEPCO. The assumption applied by TEPCO:– It is assumed that the whole molten fuel fell into
containment in Unit 1.
Sump Pit
RPV

Drywell
Floor

PCV

130ºOuter Wall

Inside Wall
Inner Skirt

ConcreteMolten Fuel
Sump Pit

http://www.tepco.co.jp/en/nu/fukushimanp/images/handouts_111130_08-e.pdf•Model for distribution of molten fuelon drywell floor
The amount of molten fuel in a sump pit isdetermined according to this distribution.
105

Ablation Profile:
It was confirmed that both results were consistentConservative

Less conservative

By TEPCO
MAAP(DECOMP): 0.65 m

By JNESCOCO Code horizontal: 0.45 m
vertical: 0.38 m

1.2m
1.45m

1.0

0.81m
0.65m

1.02m
0.65m
PCV Wall
1.02m

0.65m

0.5
Y (m)

CoriumAblation
Front

Corium

0.00.45 m

-0.5

-1.0
-1.5

0.38 m
-1.0

-0.5

0.0

0.5

1.0

1.5

X (m)

• It is difficult to evaluate the amount of molten fuel that fell
into the containment at present.
• It is expected that useful information be obtained throughR&D at the Fukushima site for further assessment of MCCI.
113. Accident Progression at Unit 3
Major Improvements• In the previous analysis, large decrease in primarypressure just after starting HPCI could not be reproduced.
– It was reported from TEPCO that HPCI injection flow rate wasreduced by using the test line. In the present analysis, this wastaken into consideration.

• Also, agreement with the plant data was poor on thecontainment pressure.
– In the present analysis, thermal stratification is presumed to takeplace at suppression chamber (S/C).
– Operation of containment spray was also considered based on theinformation from TEPCO.

126

RPV and PCV Pressures in Previous Analysis:
Results show significant differences from plant records
Unit 3Significant difference after RCIC
stopped and HPCI started

Significant
difference during
RCIC operation○○ Plant records

RCIC stopped
HPCI startedRef.: Nuclear Emergency Response Headquarters Government of Japan, "Report of Japanese Government to the IAEA Ministerial
Conference on Nuclear Safety - The Accident at TEPCO's Fukushima Nuclear Power Stations - ," June 201113

HPCI Operation Using Test Line• In the present analysis, the HPCI operation using the test
line was modeled. Large steam flow to HPCI turbine
caused rapid decrease in RPV pressure.
Steam flow toHPCI turbine

Test line

Water flow
back to CST
Water flow toRPV

http://www.tepco.co.jp/en/nu/fukushima-np/images/handouts_120312_04-e.pdf147

Major Assumptions:• In order to simulate the thermal stratification at suppression
chamber (S/C), the S/C was modeled by two volumes,
upper and lower parts, in the present MELCOR analysis.
• A CFD (computational fluid dynamics) analysis was done toestimate the temperature difference between the upper andlower part of S/C (～20K).

Temperature
distribution after 20hour operation of RCIC
Modeling in CFD analysis

Thermal Stratification• Operation of containment spray:– Flow rate of S/C and D/W spray: 50 m3/h
15

RPV and PCV Pressures in the Present analysis:Results were significantly improved
Unit 3
Venting repeatedseveral times
RPVdepressurization
just after start ofHPCI○○ Plant records
Operation ofcontainment
spray wassimulated.

Thermalstratification
is presumed
at S/C.

168

4. Hydrogen Explosions at Units 1 and 3• Analysis was done with FLUENT, a CFD code, forhydrogen transport and mixing, and AUTODYN
for structural analysis of detonation.
Major Assumptions:• In Unit 1, Hydrogen of 400 kg was released tothe top floor (5F) of reactor building (R/B) andignited there.
• In Unit 3, Hydrogen of 1000kg was released tothe first floor (1F) of R/B and ignited there.

17Hydrogen Explosion at Unit 1Analysis well reproduced theobserved explosion:• Explosion developed horizontally.
• Walls and roofs of top floor (5F)
were largely damaged and debris
scattered around R/B.

Web site: Fukushima
Central Television

Web site: TEPCO

Potential location of leakage: top head flange of PCV[m/s]ignition point[kPa]

t = 10 (ms)t = 20 (ms)
圧力コンター図(10～100ms)

t = 100 (m s)t = 100 (ms)★

Propagation of detonation pressure in Unit 1

Detonation analysis
with AUTODYN:
189

Hydrogen Explosion at Unit 3
Analysis well reproduced theobserved explosion:• Explosion developed vertically.
• Walls and roofs of top floor (5F)
were largely damaged.
• Adjacent low building was also
damaged.
• Locally high dose rate wasdetected at 1st floor.

Web site: TEPCO
Web site: Fukushima
Central Television

The top of a two-story building
next to the R/B have also been[m/s]damaged.

Potential location of leakage: device hatchignition point[kPa]

t = 5 (ms)t = 35 (ms)t = 100 (ms)★

Propagation of detonation pressure in Unit 3

Detonation analysis with AUTODYN:
195. Source Terms• The radioactive release rates obtained in thepresent analysis with MELCOR were comparedwith onsite monitoring data.
• The total amounts of releases were also
evaluated by several other organizations:– TEPCO, JAEA, JAMSTEC and CREEPI applied“inverse analysis” which evaluated the source
terms from the monitoring data andmeteorological data.

2010

Comparison between Release Rates Calculated withMELCORE and Onsite Monitoring Data：• Release timings are relatively in good agreement withthe monitoring data

1st Vent (Unit 1)

2nd Vent (Unit 1)

Vent (Unit 3)

From Unit 2

PCVleakage(Unit 1)21

Amount of Releases Evaluated in Several
Organizations:• Results are reasonably consistent with each other
Organiza Amount of Releases (PBq)
tion
I-131
Cs-134
Cs-137

Duration

250-340 8.3-15

Mar. 11-Mar. 17

JNES

7.3-13

TEPCO

500

1010

JAEA

120-9

JAMSTECCRIEPI

11Land side
only

9.7
5.5-5.7

Mar.12-May 6
Mar.21-May 6
To the ocean

3.6Mar. 26-Sep. 30
To the ocean

3.5

Inverse analysis from monitoring data
Severe accident progression analysis

Sea sideonly(1 PBq = 1015Bq)

22

11

6. Environmental Consequences
Calculated concentrations ofCs-137 in soil are in good
agreement with the measured
data

2

Calculation (Bq/m )• Very recently, JAEA conductedpreliminary environmental
consequence analysis with theOSCAAR* code developed byJAEA for level 3 PSA.
• The measured meteorological
data were used.
• The source terms calculatedby MELCOR, shown in Slide 21,were used.
*: T. Homma, et al, “Uncertainty and sensitivity
studies with the probabilistic accident consequenceassessment code OSCAAR,” Nuclear Engineering andTechnology, 37(3), 245-258 (2005).

2

Measured data (Bq/m )Correlation between measured data andcalculation (Cs-137 concentration in soil)
23

Preliminary Environmental Consequence
Analysis by JAEA/JNES
Concentration of Cs-137 in soil

Environmental consequence analysis by
JAEA (OSCAAR) using the source terms byJNES (MELCOR)

Monitoring data by MEXT
(October to Novemer, 2011)
http://radioactivity.mext.go.jp/ja/contents/5000/4901/24/1910_1216.pdf
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7. Summary• Regarding SA progression including source terms, mostof the phenomena that took place during the accident
have become reasonably well understood by efforts
made by various organizations.
• It is still difficult to predict when and how much molten
fuel fell into the containment. On this point, it isexpected that new information / data will be obtained
through R&D activities at the Fukushima site fordecommissioning.
• New attempts, such as environmental consequence
analysis with using the source terms calculated by SA
progression analysis, are expected to be able to obtainbetter understanding.
25Acronyms
AM
D/W
HPCIIRM
MCCIR/B
PCV
RCIC
RPVSA
S/C

: Accident Management: Dry Well
: High Pressure Coolant Injection: Intermediate Range Monitor: Molten Core Concrete Interaction: Reactor Building: Primary Containment Vessel: Reactor Core Isolation Cooling System: Reactor Pressure Vessel: Severe Accident: Suppression Chamber (Suppression Pool)

Organization
CRIEPI
JAEA
JAMSTECTEPCO: Central Research Institute of Electric Power Industry: Japan Atomic Energy Agency: Japan Agency for Marine-Earth Science and Technology: Tokyo Electric Power Company
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