Estimation of current status inside
RPV and PCV at Fukushima daiichi NPS
July 3, 2017
Tokyo Electric Power Company
Holdings, Incorporated
Shinya Mizokami
This document is based on the result of Subsidy Program "Project of Decommissioning
and Contaminated Water Management (Upgrading Level of Grasping State inside
Reactor)," which is a joint project by International Research Institute for Nuclear
Decommissioning (IRID) and the Institute of Applied Energy (IEA).
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1Contents

0. Background
1. Unit 1 Estimation Diagram and its Basis
2. Unit 2 Estimation Diagram and its Basis
3. Unit 3 Estimation Diagram and its Basis
4. Summary and Future Plan

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2

0-1 Background
It is important to understand the status inside RPV and PCV for thefuel debris removal.
TEPCO has been estimating it since the accident at the Fukushima
Daiichi NPS.
TEPCO is currently proceeding with the estimation in collaboration
with IRID (International Research Institute for Nuclear
Decommissioning) and IAE (Institute of Applied Energy). This
presentation shows the distribution diagrams of fuel debris in Unit
1, 2, and 3 based on the various information and the latest
knowledge obtained up to now.

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

30-2. Background (Approach for Creating Illustrations of Estimation)
Aggregated Information on each Unit
Aggregated Information on Unit 1
Reactor building (1F to 4F)

Reactor well

Operation Flow・60 mSv at 2 to 2.5 m above the shield plug (2011) [4]8・Roof collapsed in a nearly planar shape [18]・Significant contamination up to 121 mSv/h at about 1 m above
the roof directly above the reactor (8/2015) [19]・The temperature immediately above the reactor is higher thanthose in other areas (10/2011)[20]. It is estimated to be aneffect caused by heat generation of FP adhered around theshield plug・Observed an uplift on the well plugs in the upper and middle
stages of reactor [32]・ Estimated a high dose from the・Ambient dose is several mSv/h to several tens mSv/h [4]4-8
possibility of leakage at PCV top head・A high dose around RCW piping [4]4-7
・TIP room has significant contamination of l00 mSv/h or higher near pene ofcontainment vessel [5]5・The investigation of the reactor building on the first floor confirmed a high dosearound the AC system piping [6]167-176 (estimated to be an effect caused by vent)・On the fourth floor, the southwestern area is severely damaged, while damage tothe east side is minor; the ceiling in the northwest area (fifth floor) collapsed [21]The significantly damaged area is estimated to have been affected by blast anddebris caused by hydrogen explosion on the fifth floor・Leakage points exist due to・Detected Hydrogen of 10000 ppm or more from PCV spray piping systemdeteriorated sealing (steam/FP(9/2011)[27]
release)・Dose around HVH top plate and duct in MSIV room was up to 1096 mSv/h(11/2015)[15]. Upper HVH is estimated to be the dose source

PCV top head

Pressure vessel upper part・ The upper and lower parts of RPV
indicate almost the same temperature
value [9]・ A large amount of FP adhesion
・ The chemical form (water-soluble/waterinsoluble, etc.) of adhered FP is unknown.
・ The degree of reevaporation of adhered
FP is unknown
・ A leak point is estimated to have been
formed near the pressure vessel upper
part (MS piping, etc.) [29]Core・ The remaining amount is considered
to be almost 0 also from the result ofanalysis [22]・ Muon measurement performed (Febto May 2015).
No large fuel mass can be observedin the core [2]12-21
・ No water level in core area [10]RPV lower head・ Estimated that there is a heat source near
CRD piping from HVH temperature change
・ There is a correlation between the water
injection flow rate of the feed water systemand some HVH temperature behavior [11](Estimated that heat sources exist near CRD
piping in northwest, northeast, andsouthwest)
・ Checked the contact state of control rod
position detector. No clear indication that
supports estimation of the state of lower
RPV is found [23]・PCV temperatures are almost equal
to RPV temperatures, but some
PCV temperatures are higher [9]・ Estimated that damaged points exist
because water level cannot be
formed in RPV

FDW/CS piping
・ Temperature response to thechange in water injection state isgentle [9](from T/B)

PCV vent/exhaust stack・ Analysis results pointed out downward
shifting due to creep deformation anddeformed support casting.
・ Estimated a large amount of FP
adhered on separator and dryer・ Possibility of cesium taken into theoxide layer of steel material・ Possibility of cesium combined withmolybdenum, boron, and silicon

・ Water level on body side: System A
65%, System B 85% (10/18/2011)[3]・ No damage leading to outflow of
coolant to the outside equipment
or piping was observed [3]・ PCV temperatures are almost equal
to RPV temperatures, but some PCV
temperatures are higher [9]Torus room

Drywell

PCV pedestal
・ Estimated damages caused byfallen fuel debris on grating, TIP
piping, and CRD exchanger
・Water injected into RPV is
estimated to be falling on thepedestal

CS pump

DrywellCore internals

RPV lower plenum

・ Significant contamination up to 3000 mSv/h on nearside of SGTS room on the second floor (8/2011)[7]・ High contamination exceeding 10 Sv/h on the exhauststack (near SGTS piping joint) shared by Units 1 and 2(8/2011)[8]. Dose dropped to 2 Sv/h (2015/10) [26]・ Confirmed a deformed fracture on a part of diagonal(support) of the exhaust pipe common to Units 1 and 2.
Estimated it to be a damage due to hydrogen explosion[26]
・ Unit 2 SGTS contamination of several Sv/h is estimatedto be derived from Unit 1 because the rupture disk of
Unit 2 is not damaged
During the investigation of pedestal-side first floor upper grating, despite the・ Detected Cs134 (826 kBq/cm3) Csl37 (51.9 kBq/cm3) from presence of fallouts, no significant damage was observed on the structure.
accumulated water in pit (9/12/2016) [33]Average dose is several Sv/h. Confirmed presence of something like pipe
shielding body around PLR piping [1](Estimated melted lead mat)・Temperature response to the changes in water injection state is gentle.
Some PCV thermometer indications have been observed to increase due to theeffect of measures such as nitrogen sealing without water injection. [9]・Relative to the assumed decay heat, the observed temperature is low ・From the pressure, some leakage is estimated to be in the gas phase・Filled the water level gauge pipe with water [10] (assuming that the waterlevel gauge pipe is sound)・ White powdery deposits (sampled)around Hpa steam pipe pene X-53
(bellows cover, floor, and wall
surface). Leakage marks are seen
between HPCI piping and bellows
cover as well as at the base of bellows
cover and biological shielding wall.
The highest dose around the base (upto 7 Sv/h) [l5]・ Measurement result/observation information
available
・ Estimation from measurement result/observation
information
・ Accident analysis or qualitative estimation
・ Pick

PCV pedestal floor・ Suppressed the water level to no more than
O.P. 3200 to prevent outflow to the turbine・
building• Vacuum break line bellows. Leakage
・observed from sand cushion drain pipe [12][13]・・ 200 to 2400 mSv/h on catwalk (5/2014)[16]・ There should be some communication port・
between the reactor and turbine buildings[25]• Significant contamination up to 920 mSv/hin the air and 800 mSv/h underwater [30]MS piping

Turbine building・Ambient dose is several uSv/h toseveral hundreds mSv/h [4]9-11
Underground floor has high dose[4]9Significant contamination up to 5000
mSv/h on near side of SGTS room on
the second floor (5/2011) [7]
Estimated that high concentration
contaminated water flowed fromvacuum break pipe bellows and sandcushion・ The cause of high dose observed near RCW piping
in the reactor building is estimated to be the debris
damaging equipment drain sump in the pedestal,・ S/C with nearly full water (gas generated in the early stage of the accident
resulted in radioactive materials contaminating
remains slightly under nitrogen sealing) [17]RCW system. Therefore, debris are estimated to
S/C is estimated to be almost soundhave fallen in the pedestal [28]Fallen debris are estimated to have eroded the・ From the result of the torus room dose investigation, a high Cs concentration waspedestal floor and sump to a certain extent [22]estimated on the building walls and S/C walls or in accumulated water in S/C [24]Pressure control room

・Water level about 3 m (similar level of water; leakage from vacuum breaktube bellows is observed) [14] 18-29
・Camera investigation found deposits on the floor [14] 18-29. Second
investigation found an increasing tendency [31]. Estimated that it is not a high･dose because camera noise is small
・Analysis shows that debris diffused to D/W floor and reached PCV shell [22]Aggregated various information leading to estimationof fuel debris/FP distribution in each unit

Illustration of Estimated Fuel Debris Distribution
LegendIllustration of Estimated Debris Distribution and RPV/PCV State of Unit 1
・Possibility of forming molten pool in thereactor during accident (generalestimation)･Possibility of shroud breakage (generalestimation)
・If the shroud is broken, there is apossibility of breakage of the jet pump
resulting from intrusion of molten fuel into
the downcomer (general estimation)

・Estimated that most of fuel melted andno fuel rod remains based on the muon
measurement, analysis result, and the factthat a water level is not formed (actual
measurement and analysis)
・Abundance of debris is estimated to beslight from the cooling achieved before
starting CS injection (12/10/2011)
・Estimated to be general oxide debris
solidified with molten fuel (generalestimation)･If heat transfer from high temperature
fuel debris is small, CRGT can beunmelted and remains (generalestimation)

・Possibility of accumulation in stagnant
spot when there are particulate debris(general estimation)

・Damaged points exist on the lower
plenum because water level cannot be
formed up to the core (estimation basedon actual measurement)･Possibility that bottom drain at thebottom of lower plenum is fragile anddamaged (general estimation)･Possibility that fuel fallen in the lower
plenum remains at the bottom of RPV(general estimation)

・Lower pedestal walls nearsump can be partially eroded by
MCCI (Generalestimation/analysis)
・The result image taken with theunderwater CCD camera on D/W floor
shows something like deposited
sediments (actual measurement)

・If water is pooled on PCV floor, particulate
debris will be formed.
・Possibility of accumulation in stagnant
spot when there are particulate debris(general estimation)･Fuel debris that caused MCCI is mixed
with concrete.
(General estimation/analysis)
・Estimated that RCW piping inequipment drain sump was damaged andradioactive materials contaminated RCW
system (estimation based on actual
measurement)

・From HVH temperature, it is estimatedthat debris exists near CRD (estimation
based on actual measurement andanalysis)・Because a specific HVH thermometershows high temperature rise when the
FDW flow rate is decreased, debris can
exist near CRD on the outer periphery(no distinction between adhesion toouter surface and intrusion into inside)or RPV can be damaged directly above it(estimation based on actual
measurement)
・Possibility of part of fuel debris
solidified without causing MCCI
(general estimation)･Possibility of diffusion of debris to theD/W floor through the pedestal
opening (general estimation/analysis)
・Estimated that PCV had a damage
due to water leakage from sandcushion drain pipe (actual
measurement and analysis)

Oxide debris (porous)

Granular debris
Concrete mixed debris

Sound CRGT
Damaged CRGT
Sound CRD
CRD (debris inside)
Sound shroud
Sediment (unknown material)

RPV damaged point
Burning fuel*
Fuel rod*
Oxide debris*
Heavy metal debris*
Pellets*
Powder pellets*
Clad residue*
Molten core internals*

Solidified B4C*
Control rod mixture melts**Not used in Illustration of Estimated Unit 1

The analysis result has still uncertain/unknown points.
Consider:・Measured data during accident
・Information obtained from cooling of fuel debris after accident
・Information obtained from internal investigation
for comprehensive estimation of fuel debris distribution

TEPCO has been discussing this fuel debris distribution with domestic and foreign experts from the meanings
of information while transmitting observational information.
TEPCO will continue to upgrade the distribution diagram based on new findings.
⇒The following slides show the basis of representative estimation.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

40-3. Background （Overview of RPV and PCV（Unit 2））
５Ｆ

Shield plug
RPV

４Ｆ
３Ｆ
Core region
２Ｆ

Ground level（About 10m above
the sea level）

CS

FDW
PCV（D/W）

X-53 penetration

１Ｆ

X-6 penetration

Access port

PCV（S/C）

Direction
for sump

CRD exchange rail

Inside pedestal

Direction
for accessport©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Grating50-2. Background (Approach for Creating Illustrations of Estimation)
Aggregated Information on each Unit
Aggregated Information on Unit 1
Reactor building (1F to 4F)

Reactor well

Operation Flow・60 mSv at 2 to 2.5 m above the shield plug (2011) [4]8・Roof collapsed in a nearly planar shape [18]・Significant contamination up to 121 mSv/h at about 1 m above
the roof directly above the reactor (8/2015) [19]・The temperature immediately above the reactor is higher thanthose in other areas (10/2011)[20]. It is estimated to be aneffect caused by heat generation of FP adhered around theshield plug・Observed an uplift on the well plugs in the upper and middle
stages of reactor [32]・ Estimated a high dose from the・Ambient dose is several mSv/h to several tens mSv/h [4]4-8
possibility of leakage at PCV top head・A high dose around RCW piping [4]4-7
・TIP room has significant contamination of l00 mSv/h or higher near pene ofcontainment vessel [5]5・The investigation of the reactor building on the first floor confirmed a high dosearound the AC system piping [6]167-176 (estimated to be an effect caused by vent)・On the fourth floor, the southwestern area is severely damaged, while damage tothe east side is minor; the ceiling in the northwest area (fifth floor) collapsed [21]The significantly damaged area is estimated to have been affected by blast anddebris caused by hydrogen explosion on the fifth floor・Leakage points exist due to・Detected Hydrogen of 10000 ppm or more from PCV spray piping systemdeteriorated sealing (steam/FP(9/2011)[27]
release)・Dose around HVH top plate and duct in MSIV room was up to 1096 mSv/h(11/2015)[15]. Upper HVH is estimated to be the dose source

PCV top head

Pressure vessel upper part・ The upper and lower parts of RPV
indicate almost the same temperature
value [9]・ A large amount of FP adhesion
・ The chemical form (water-soluble/waterinsoluble, etc.) of adhered FP is unknown.
・ The degree of reevaporation of adhered
FP is unknown
・ A leak point is estimated to have been
formed near the pressure vessel upper
part (MS piping, etc.) [29]Core・ The remaining amount is considered
to be almost 0 also from the result ofanalysis [22]・ Muon measurement performed (Febto May 2015).
No large fuel mass can be observedin the core [2]12-21
・ No water level in core area [10]RPV lower head・ Estimated that there is a heat source near
CRD piping from HVH temperature change
・ There is a correlation between the water
injection flow rate of the feed water systemand some HVH temperature behavior [11](Estimated that heat sources exist near CRD
piping in northwest, northeast, andsouthwest)
・ Checked the contact state of control rod
position detector. No clear indication that
supports estimation of the state of lower
RPV is found [23]・PCV temperatures are almost equal
to RPV temperatures, but some
PCV temperatures are higher [9]・ Estimated that damaged points exist
because water level cannot be
formed in RPV

FDW/CS piping
・ Temperature response to thechange in water injection state isgentle [9](from T/B)

PCV vent/exhaust stack・ Analysis results pointed out downward
shifting due to creep deformation anddeformed support casting.
・ Estimated a large amount of FP
adhered on separator and dryer・ Possibility of cesium taken into theoxide layer of steel material・ Possibility of cesium combined withmolybdenum, boron, and silicon

・ Water level on body side: System A
65%, System B 85% (10/18/2011)[3]・ No damage leading to outflow of
coolant to the outside equipment
or piping was observed [3]・ PCV temperatures are almost equal
to RPV temperatures, but some PCV
temperatures are higher [9]Torus room

Drywell

PCV pedestal
・ Estimated damages caused byfallen fuel debris on grating, TIP
piping, and CRD exchanger
・Water injected into RPV is
estimated to be falling on thepedestal

CS pump

DrywellCore internals

RPV lower plenum

・ Significant contamination up to 3000 mSv/h on nearside of SGTS room on the second floor (8/2011)[7]・ High contamination exceeding 10 Sv/h on the exhauststack (near SGTS piping joint) shared by Units 1 and 2(8/2011)[8]. Dose dropped to 2 Sv/h (2015/10) [26]・ Confirmed a deformed fracture on a part of diagonal(support) of the exhaust pipe common to Units 1 and 2.
Estimated it to be a damage due to hydrogen explosion[26]
・ Unit 2 SGTS contamination of several Sv/h is estimatedto be derived from Unit 1 because the rupture disk of
Unit 2 is not damaged
During the investigation of pedestal-side first floor upper grating, despite the・ Detected Cs134 (826 kBq/cm3) Csl37 (51.9 kBq/cm3) from presence of fallouts, no significant damage was observed on the structure.
accumulated water in pit (9/12/2016) [33]Average dose is several Sv/h. Confirmed presence of something like pipe
shielding body around PLR piping [1](Estimated melted lead mat)・Temperature response to the changes in water injection state is gentle.
Some PCV thermometer indications have been observed to increase due to theeffect of measures such as nitrogen sealing without water injection. [9]・Relative to the assumed decay heat, the observed temperature is low ・From the pressure, some leakage is estimated to be in the gas phase・Filled the water level gauge pipe with water [10] (assuming that the waterlevel gauge pipe is sound)・ White powdery deposits (sampled)around Hpa steam pipe pene X-53
(bellows cover, floor, and wall
surface). Leakage marks are seen
between HPCI piping and bellows
cover as well as at the base of bellows
cover and biological shielding wall.
The highest dose around the base (upto 7 Sv/h) [l5]・ Measurement result/observation information
available
・ Estimation from measurement result/observation
information
・ Accident analysis or qualitative estimation
・ Pick

PCV pedestal floor・ Suppressed the water level to no more than
O.P. 3200 to prevent outflow to the turbine・
building• Vacuum break line bellows. Leakage
・observed from sand cushion drain pipe [12][13]・・ 200 to 2400 mSv/h on catwalk (5/2014)[16]・ There should be some communication port・
between the reactor and turbine buildings[25]• Significant contamination up to 920 mSv/hin the air and 800 mSv/h underwater [30]MS piping

Turbine building・Ambient dose is several uSv/h toseveral hundreds mSv/h [4]9-11
Underground floor has high dose[4]9Significant contamination up to 5000
mSv/h on near side of SGTS room on
the second floor (5/2011) [7]
Estimated that high concentration
contaminated water flowed fromvacuum break pipe bellows and sandcushion・ The cause of high dose observed near RCW piping
in the reactor building is estimated to be the debris
damaging equipment drain sump in the pedestal,・ S/C with nearly full water (gas generated in the early stage of the accident
resulted in radioactive materials contaminating
remains slightly under nitrogen sealing) [17]RCW system. Therefore, debris are estimated to
S/C is estimated to be almost soundhave fallen in the pedestal [28]Fallen debris are estimated to have eroded the・ From the result of the torus room dose investigation, a high Cs concentration waspedestal floor and sump to a certain extent [22]estimated on the building walls and S/C walls or in accumulated water in S/C [24]Pressure control room

・Water level about 3 m (similar level of water; leakage from vacuum breaktube bellows is observed) [14] 18-29
・Camera investigation found deposits on the floor [14] 18-29. Second
investigation found an increasing tendency [31]. Estimated that it is not a high･dose because camera noise is small
・Analysis shows that debris diffused to D/W floor and reached PCV shell [22]Aggregated various information leading to estimationof fuel debris/FP distribution in each unit

Illustration of Estimated Fuel Debris Distribution
LegendIllustration of Estimated Debris Distribution and RPV/PCV State of Unit 1
・Possibility of forming molten pool in thereactor during accident (generalestimation)･Possibility of shroud breakage (generalestimation)
・If the shroud is broken, there is apossibility of breakage of the jet pump
resulting from intrusion of molten fuel into
the downcomer (general estimation)

・Estimated that most of fuel melted andno fuel rod remains based on the muon
measurement, analysis result, and the factthat a water level is not formed (actual
measurement and analysis)
・Abundance of debris is estimated to beslight from the cooling achieved before
starting CS injection (12/10/2011)
・Estimated to be general oxide debris
solidified with molten fuel (generalestimation)･If heat transfer from high temperature
fuel debris is small, CRGT can beunmelted and remains (generalestimation)

・Possibility of accumulation in stagnant
spot when there are particulate debris(general estimation)

・Damaged points exist on the lower
plenum because water level cannot be
formed up to the core (estimation basedon actual measurement)･Possibility that bottom drain at thebottom of lower plenum is fragile anddamaged (general estimation)･Possibility that fuel fallen in the lower
plenum remains at the bottom of RPV(general estimation)

・Lower pedestal walls nearsump can be partially eroded by
MCCI (Generalestimation/analysis)
・The result image taken with theunderwater CCD camera on D/W floor
shows something like deposited
sediments (actual measurement)

・If water is pooled on PCV floor, particulate
debris will be formed.
・Possibility of accumulation in stagnant
spot when there are particulate debris(general estimation)･Fuel debris that caused MCCI is mixed
with concrete.
(General estimation/analysis)
・Estimated that RCW piping inequipment drain sump was damaged andradioactive materials contaminated RCW
system (estimation based on actual
measurement)

・From HVH temperature, it is estimatedthat debris exists near CRD (estimation
based on actual measurement andanalysis)・Because a specific HVH thermometershows high temperature rise when the
FDW flow rate is decreased, debris can
exist near CRD on the outer periphery(no distinction between adhesion toouter surface and intrusion into inside)or RPV can be damaged directly above it(estimation based on actual
measurement)
・Possibility of part of fuel debris
solidified without causing MCCI
(general estimation)･Possibility of diffusion of debris to theD/W floor through the pedestal
opening (general estimation/analysis)
・Estimated that PCV had a damage
due to water leakage from sandcushion drain pipe (actual
measurement and analysis)

Oxide debris (porous)

Granular debris
Concrete mixed debris

Sound CRGT
Damaged CRGT
Sound CRD
CRD (debris inside)
Sound shroud
Sediment (unknown material)

RPV damaged point
Burning fuel*
Fuel rod*
Oxide debris*
Heavy metal debris*
Pellets*
Powder pellets*
Clad residue*
Molten core internals*

Solidified B4C*
Control rod mixture melts**Not used in Illustration of Estimated Unit 1

The analysis result has still uncertain/unknown points.
Consider:・Measured data during accident
・Information obtained from cooling of fuel debris after accident
・Information obtained from internal investigation
for comprehensive estimation of fuel debris distribution

TEPCO has been discussing this fuel debris distribution with domestic and foreign experts from the meanings
of information while transmitting observational information.
TEPCO will continue to upgrade the distribution diagram based on new findings.
⇒The following slides show the basis of representative estimation.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

6

1. Unit 1 Estimation Diagram and its Basis©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

1-1 Decay Heat of Unit 1 and Heat
Generated by Water-Zr Reaction
100

Period without

No water
injectionwater injection

90

Unit 2

7

Unit 3

Decay
heat
Decay
heat by ORIGEN
Heat
generatedby water-Zr(MAAPbyevaluation)
Enthalpy
from
W-Zrreaction
reaction
MAAP

Generated Heat [MW]80

water level

Water level drop
decreasedue
to decay by
decayheat
heat

70
60
50

Water
Water injection
injection by fire engines started
by fire truck

40
30
20

IC
Cooling

core
damage,Core damage
andcore
meltduebyto
core melt
water-Zr reaction
water-Zr
reaction

10
0
3/11
12:00

3/12
0:00

3/12
12:00

3/13
0:00

3/13
12:00

Unit 1 became uncoolable immediately afterDate/time
the tsunami arrival, which resulted in core damage/core
melt in March 11 2011. Insufficient fire engine water injection made no contribution to water-Zr
reactions©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1-2 Unit 1 Estimation Diagram

Unit 2･If heat transfer from high temperature
fuel debris is small, CRGT can beunmelted but remains
・Damaged points exist on the lower plenum
because of failing to form a water level up to
the core･Possibility that bottom drain at the bottomof lower plenum is fragile and damaged･Possibility that fuel fallen in the lower
plenum remains at the bottom of RPV・Lower pedestal walls near sump can bepartially eroded by MCCI・If water is pooled on PCV floor,
particulate debris will be formed.
・Possibility of accumulation instagnant spot when there areparticulate debris

・Fuel debris that caused MCCI is mixed withconcrete.
・Estimated that RCW piping in equipment
drain sump was damaged and radioactive
materials contaminated RCW system

Unit 3

Legend・Estimated that most of fuel melted
and almost no fuel rod remains basedon the muon measurements, analysis
results, and the fact that a water levelis not formed
・Estimated almost no heat source
remained in core region from the factthat sub-cooling conditions wasachieved before starting CS injection(12/10/2011)
・Estimated to be oxide debris
solidified with molten fuel・Possibility of forming molten pool in thereactor during accident･Possibility of shroud breakage
・If the shroud is broken, there is apossibility of breakage of the jet pump
resulting from intrusion of molten fuelinto the downcomer

・The video image taken by the underwater
CCD camera on D/W floor shows something
like deposited sediments

Unit 1

8

Oxide debris (porous)
Granular debris
Concrete mixed debris
Sound CRGT
Damaged CRGT
Sound CRD

・Possibility of accumulation instagnant spot when there areparticulate debris

CRD (debris inside)

・From HVH temperature, estimatedthat debris exist near CRD
・Because a specific HVH thermometershows high temperature rise when the
FDW flow rate is decreased, debris can
exist near CRD on the outer periphery(no distinction between adhesion toouter surface and intrusion into inside)or RPV can be damaged directly aboveit

Sediment (unknown material)・Possibility of part of fuel debris
solidified without causing MCCI･Possibility of diffusion of debris
to the D/W floor through thepedestal opening

Sound shroud
RPV damaged point
Burning fuel*
Fuel rod*
Oxide debris*
Heavy metal debris*
Pellets*
Powder pellets*
Clad residue*
Molten core internals*

・Estimated PCV had a damage
based on observed water leakage
from sand cushion drain pipe

About 2.6m
Concrete layer：About 7.6m

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Solidified B4C*
Control rod mixture melts**Not used in Illustration of Estimated Unit 1

1-3 Estimated State of Core

Unit 1

Unit 2

Unit 3

・Estimated that most of fuel melted andalmost no fuel rod remains based on themuon measurement, analysis result, andthe fact that a water level is not formed
・Estimated almost no heat source
remained in core region from the fact that
sub-cooling conditions was achieved before
starting CS injection (12/10/2011)©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

9

1-4

Temperature Behavior before Starting
CS Water Injection

Unit 1[Considerations]With regard to measuring instruments, some of them can be beyond their operating
environment conditions due to earthquakes, and the subsequent event evolution and others
can provide incorrect readouts. In order to grasp the plant status, taking into considerationsuch uncertainty of instruments, a comprehensive judgment is made focusing on the trend ofchanges using information obtained from multiple instruments.

Unit 2

Unit 310
VESSEL flange
Vessel flange
RPV bellow seal
(HVH-12A)

12/10
CS water
injection
startedFeed water
nozzle N4B(terminal)
vessel core
Safety valve
exhaust203-4A (1)Pressure vessel
Lower Part
(Lower head)
D/W HVH return(HVH-12C)
CRD upper housing
CRD lower housing

100

S/C pool watertemperature A
S/C pool watertemperature B

Below 100 ℃
For Unit 1, water injection from CS system started on December 10, 2011. Before that,
however, most of the observation points in the containment vessel fell below 100 ℃.
Therefore, estimated that almost no fuel remains in the core.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1-5 Result of Muon Measurements

Image of estimated muon measurement result based on design
drawing (Area enclosed by a broken line shows the core)

Unit 1

Unit 2

Unit 3

11

Image of muon measurement using 96 days’ worth of data
(Area enclosed by a broken line shows the core)

High-absorbent objects (except for fuel) such as pressure vessel walls,
containment vessel walls, and reactor building walls can be confirmed bymuon measurements. (Measurement period: February 9 to May 21, 2015)
Meanwhile, no large fuel mass exceeded 1 m in the core （resolution of themeasurement）.
⇒Estimated that almost no fuel remains in the core.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1-6 Estimation of Pedestal Part

Unit 1

Unit 2

Unit 3

12・Estimated that RCW
piping in equipment drainsump was damaged andradioactive materials
contaminated RCW
systemAbout 2.6m

Concrete layer：About 7.6m

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

1-7 Contamination of RCW System

First floor

Second floor

Unit 2

Unit 3

13

Ambient Dose Rate of Unit 1 Reactor Building(Unit: mSv/h)

Third floor

MG Set A
Hundreds mSv/hFourth floor

RCW heat exchanger>1000 mSv/h

RCW surge tank
90 mSv/h

DHC (Drywell humidity control)
120 mSv/h

RCW system is a closed loop* and unlikely to experience high dose contamination innormal conditions. However, high dose was observed around equipment to be cooled by
RCW over multiple floors in the reactor building.
*Closed loop: Piping network without opening in PCV©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

1-8 Contamination of RCW System

Unit 2

Unit 3

14

Reactor building fourth floor
90 mSv/h
Reactor building first floor
Around RCW system piping
Hundreds mSv/h

ＲＣＷ
RCW

surge
サージ
tank
タンク

To other

他機器へ
device(1)
Containment vessel

ＲＣＷ
RCW

Pedestal(2)

床ドレン
Floor drainsump pit
サンプピット

RCW
heat
ＲＣＷ
exchanger
熱交換器

pump
(3) ポンプ

RCW system
機器ドレン
Equipment
drain
sump pit
サンプピットReactor building second floor
RCW pump
Around RCW heat exchanger
Over 1000 mSv/h

Because RCW system is cooling the equipment drain sump in the pedestal, the contaminants in
RCW system are estimated to be:(1) Molten fuel fell into the equipment drain sump located in the containment vessel pedestal⇒ (2) Molten fuel damaged RCW piping in equipment drain sump
⇒ (3) In Unit 1, with the containment pressure increased significantly during the accident
progression, water and steam containing radioactive materials transferred to RCW piping©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

1-9 Estimation of Pedestal/Drywell Areas？

About 2.6m

Concrete layer：About 7.6m？

Unit 2

Unit 315

・Estimated that PCV hada damage based on
observed water leakage
from sand cushion drain
pipe©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1-10 Leakage from Sand Cushion Drain
Pipe

Unit 1

Unit 2

Unit 3

16

Using the video camera mounted on the remote boat, confirmed water leakage from the vent pipeand sand cushion drain pipe (November 13 and 14, 2013)
Vacuum break line

(1) Leakage from sand cushion drain pipe
Sand cushion
Sand cushion drain pipe

Point A: Vent pipe sleeve
Point B: Vent pipe
Point C: Sand cushion drain pipe
Point C: Sand cushion ring header

Drain funnel

Sand cushion drain
pipe

Point where sand cushion
drain pipe leakage is observed
Sand cushion ring header
broken point

Sand cushion ringheader

Leakage is observed from sandcushion drain pipe.
Although the location cannot be
identified, the iron shell below thewater surface could possibly have aleakage point.

To floor funnel©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1-11 D/W Floor Condition

Unit 1

Unit 2

Unit 3

17

・The video image taken by
the underwater CCD
camera on D/W floor
shows something likedeposited sediments

About 2.6m

Concrete layer：About 7.6m

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

1-12 Result of PCV Internal Investigation

About 60°

X100B pene
About 800

Unit [mm]Unit 2

18

90°

X100B penetration

PCV shell

180°

0°Grating OP
9708

As of 10/10 11:00 a.m.

CCD camera
dosimeter

About
3500

190°

Expected water level

Bottom of D/W

Unit 3

Pedestal wall
270°

The dose in the air on the first floor in PCV is 4.7 to 9.8 Sv/h
About 11.1 Sv/h at the end of pene when inserting the instrument
Dose distribution tends to decrease as height decreases
0.5 Sv/h underwater in accumulated water

As a PCV internal investigation, CCD camera was inserted from X100B pene (October 11, 2012)©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1-13 Result of PCV Internal Investigation

Containment vessel bottom (indent) [close-up]

Unit 1

Unit 2

Unit 319

Containment vessel bottom (surface) [close-up]

・There are bluish fragments
・Deposits on the bottom

There are some deposits having a certain thickness at the bottom ofcontainment vessel. Deposit seems not to be hard solids as it had a dent when
the camera bumped it. Some deposits are bluish fragments. (Possibility of lead)©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

1-9 Estimation of Pedestal/Drywell Areas？

About 2.6m

Concrete layer：About 7.6m？

Unit 2

Unit 320

・Estimated that PCV hada damage based on
observed water leakage
from sand cushion drain
pipe©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

1-14 Result of PCV Internal Investigation

Unit 1

Unit 2

Unit 321

In March 2017, the PCV internal investigation was conducted to check the diffusion
condition of fuel debris in the underground level out of pedestal and to check whether or
not the fuel debris reached the PCV shell.
Measurement points of the investigation are as
follows:

Measurement
point

DO

Pedestal

Opening
CRD rail

Estimation
Diffusion of fuel debris from the drain sump

BG

Background levels against D0-D3 measurements

D1, D2

Diffusion of fuel debris from the opening

D3

Possibility of fuel debris reaching to PCV shell・Hang the measuring unit down to the bottom and thenmeasure the dose while pulling it up at 5-cm intervals

Primary containment
vessel

X -100B pene

Above grating

Drain sump
PCV first floor grating plan view

Lowest point

Measurement Image©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

The height from the floor
that can be measured
differs depending on thepresence or absence ofobstruction

1-15 Result of PCV Internal Investigation

Unit 1

Unit 2

Typical images taken are as follows:

Unit 322

Shadow caused byapproaching measurement
unit

Opening

Pedestal
Existing structure

3/22 D2 (3) Image near the lowest point

3/21 D3 Image near the lowest point

3/21 D2 (2) Image near the lowest point

3/21 D1 (2) Image near the lowest point

Fallout

First floor PCV
cross-section
Fallout

・Detailed measuring points are
currently under evaluation

3/20 D2 (1) Image near the lowest
point

3/21 D1 (1) Image near the lowest point

Deposits having a certain thickness are observed in each observation point in the same
way as the investigation in 2012.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

23

2. Unit 2 Estimation Diagram and its Basis©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1
2-1 Decay Heat of Unit 2 and Heat
Generated by Water-Zr Reaction

150
140
130

Decay heat
Decay
heat
Heat generated
by water-Zr
reaction
Enthalpy
from Water
Zr reaction

Generated Heat [MW]

120
100
90
80
60
40
20

Core damage
andcore
damage,core melt due tocoremelt bywater-Zr reaction
W-Zr reaction

Drawdown
water
water of
level
level
due to flashingdecreaseby
Steam cooling

50

24

Water
injection
Water
injectionbyby
fire
enginesfire
truck
startedDepressurization
SRV
manual open

70

Unit 3

Periodwithout
water injectionNowater
injectionwaterlevel
level
Water
decreasedrop
due toby
decayheat
decay
heat

110

30

Unit 2decayheat(Level drops below BAF)

RCIC cooling

10
0
3/11
12:00

3/12
0:00

3/12
12:00

3/13
0:00

3/13
12:00

3/14
0:00

3/14
12:00

3/15
0:00

3/15
12:00

At Unit 2, fuel is fully exposed because
water level dropped due to flashing.
Date/time
After that, the water-Zr reaction induced by fire engine water injection resulted incore damage and core melt.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Pressure Increase after Reactor depressurization

At Fukushima Daiichi Unit 2, the reactor
depressurization was successful upon
forced SRV opening at 6:00 p.m. on March
14, but the reactor experienced a pressure
rise three times during the period fromthat night to the next morning.
This behavior was thought to be related to
SRV opening operation (pressure increase= SRV closed, pressure decrease = SRVopen). However, this does not directly
show the open/close state of SRV.
Along with this pressure increase, anincrease in the containment vesselpressure associated with hydrogen has
been confirmed, and this pressure
increase is greatly related to the accident
progression in Unit 2.

Unit 225

Unit 3

Reactor depressurization
Reactor pressure Rises
4.0
Reactor pressure
原子炉圧力
(測定値)
(measuredvalue)

3.5
3.0
Reactorpressure [MPa(abs)]
原子炉圧力

2-2

Unit 1

2.5
2.0
1.5
1.0
0.5
0.0
3/14
18:00

3/14
21:00

3/15
0:00

3/15
3:00

Date/time

Understanding this behavior isextremely important in understanding
the accident progression and corestatus of Unit 2©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

26

Unit 3Pressure Increase after Reactor depressurization
4
RPV pressure (measured3
圧力 (MPa[abs])
Pressure

2-3

Unit 2

RPV圧力（測定値）
value)
D/W pressure (measured
DW圧力（測定値）
value)
RPV pressure (calculated
RPV圧力（計算値）
value)
D/W pressure (calculated
D/W圧力（計算値）
value)

S/C pressure (calculated
S/C圧力（計算値）

2

value)

1

0
3/14
18:00

3/14
21:00

3/15
0:00

3/15
3:00

日時by GOTHIC (Generation of Thermal HydraulicAdjusting the generation amount of steam and hydrogen
Information in Containment) that is the general flow analysis code successfully reproduced the RPV pressure
and PCV pressure.
Assumed semi-opening (40%) of SRV to reproduce the RPV/PCV pressure behavior at the second peak.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Pressure Increase after Reactor depressurization
10
Steam generation amount
蒸気発成量（左軸）

蒸気発生量
Steam generation
amount (kg/s)

8

Second peak(left axis)
Hydrogen generation
水素発生量（右軸）
amount (right axis)

Unit 227

Unit 3

0.5
Third peak

0.4First peak

6

0.3

4

0.2

2

0.1

0
3/14
18:00

3/14
21:00

3/15
0:00

0.0
3/15
3:00

水素発生量
Hydrogen generation
amount (kg/s)

2-4

Unit 1

The calculated result indicates that almost all of zirconium in the reactor reacted until the second
peak.
Estimated that most of fuel melted during the second peak from the relationship betweenhydrogen generation and energy generation due to water-zirconium reaction.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-5

Unit 2 Estimation Diagram

・Assuming the energy amount from PCV
pressure increased due to hydrogengeneration, most of fuels are estimatedto have melted
・Because of the temperature drop
confirmed during CS injection, thepresence of fuel is estimated to be on
the outer periphery of core where water
is applied by CS injection at low flow rate
(debris cannot be specifically locatedbecause the same behavior is showneven when molten fuel dropped andsolidified in a fuel support casting or
CRGT)・The result of muon measurementshows the possibility that fuel exists on
the outer periphery of core・Fuel rods can exist only on the outer
periphery if any
・Estimated to be general oxide debris
solidified with molten fuel・A shadow of the high density material that is
considered to be fuel debris has been
confirmed at the bottom of pressure vessel bymuon measurements. Possibility that fuelfallen in the lower plenum remains at thebottom of RPV・Possibility that debris intrude in CRD
accompanying breakage of CRGT・If water is pooled on PCV floor,
particulate debris will be formed.
・Possibility of accumulation in stagnant
spot when there are particulate debris

Unit 1

Unit 2・Possibility of water level formed outside
the shroud because the PLR systempressure increased when the FDW flowrate increased
・Estimated that there is no large-scale
damage of shroud from a temperaturedrop by CS injection and water level riseoutside the shroud when injection rateincreases
・Possibility of pellets remain on the outer
periphery because the fuel temperaturerise on the outer peripheral may not beso high･If heat transfer from high temperature
fuel debris is small, CRGT will be
unmelted but remains
・Possibility of accumulation in stagnant
spot when there are particulate debris
and pellets

・Possibility of part of fuel debris
solidified without causing MCCI・Steam rising is observed during PCVinternal investigation.

・CRD on the outer periphery wasconfirmed during PCV internal
investigation, and the holes on grating
panel and some melting material wereobserved which indicate RPV bottom
failure (not large, center and off-center
position)・Some of debris that fell out of the holeare estimated to adhere on CRD・Fuel debris that caused MCCI is mixed
with concrete
・MCCI is estimated to be limited becauseof no symptom of PCV shell breakage (noleakage from sand cushion drain pipe)

Unit 3

Legend28

Fuel rod
Oxide debris (porous)
Granular debris
Concrete mixed debris
Sound CRGT
Damaged CRGT
Sound CRD
CRD (debris inside)
Sound shroud
Pellets
Burning fuel*Oxide debris*
Heavy metal debris*
Powder pellets*
Clad residue*
Molten core internals*
Solidified B4C*
Control rod mixture melts**Not used in Illustration of Estimated Unit 2©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-6 Estimated State of Core
(Soundness of Shroud)

Unit 1

Unit 2

Unit 329

RPV
ShroudPLR piping
Pressuregauge

・Possibility of water level
formed outside the shroud
because the PLR systempressure increased when the
FDW flow rate increased
・Estimated that there is nolarge-scale damage of shroud
from a temperature drop by CS
injection and water level riseoutside the shroud when
injection rate increases©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-7 Water Level in Annulus Expected from PLR
Inlet Pressure when FDW Flow Rate Changes

Unit 1

Unit 2

Unit 330

Estimated water level [OP]Estimated water level from PLR pump (A)inlet pressure
Estimated water level from PLR pump (B)inlet pressure
FDW line discharge flow rate

Figure Relationship between Water Injection Volume from FDW System and Water Level in Annulus(December 2011 to February 2012)

As the water injection rate increases, the water level in annulus rises.
⇒ (two possibilities)
・Possibility of water level formed outside the shroud with damage on shroud to small extent
・Because the total water injection rate is increasing, the water level in the reactor pressure vessel canbe rising even though the shroud is damaged©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-8

Temperature Behavior when Starting CS
Water Injection

PCV

℃

300
RPV

Unit 1

Unit 2

Unit 331

Temperature Change at Each Location inside Unit 2(2011)
２号機 格納容器内部各箇所の温度変化
Containment
Vessel (2011)

RPV bellows
A: 圧力容器ﾍﾞﾛｰｼｰﾙ

250
B:Feed water
B：給水ノズル
nozzle200

SRV
C: 逃し安全弁排気150

RPV bottom
D: 圧力容器底部

100
100

CRD top
E: CRD上部

9/14

Temperature dropped

50
In-PCV
F: 格納容器内

0
3/22

4/21

5/21

6/20

7/20

8/19

9/18 10/18 11/17

Along with water injection from CS system started on September 14, 2011, thetemperatures of A to F decreased similarly⇒
The heat source is considered to be located in the area that is cooled by water injection
from CS system. This also supports that the shroud has only a minor damage.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

Unit 2

Unit 3

2-9 Estimated State of Core (Outer Periphery)•Possibility of pellets remain on theouter periphery because the fueltemperature rise on the outer
peripheral may not be so high©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

32

Unit 1

2-10

Unit 2

Unit 3Result of Simulated Fuel Assembly
Breakage Test
In order to know more about fuel behaviors such as melting and transfer in BWR
system, a test was conducted by heating the test piece that simulates the core andlower support structure of BWR system with a plasma torch.

33

Heat theupper part toover 2600 K

Placed a channel box andsimulated 48 fuel rods
(ZrO2) on both sides of thecontrol rod blade

Test
vesselAlignment with actual
reactorGas
permeables

Fuel is hard
to collapse

The test result was, partly affected byradiation, that the fuel rods
maintained their shape to a certain
degree after heating.
On the outer periphery of core whereradiation in the shroud direction islarge and the fuel temperature rise is
not so high, pellets or the like could
possibly remain.

Melts such as control
rod blades flow down

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

Unit 2

Unit 3

2-11 Estimation of Pressure Vessel Lower Part・A shadow of the high density
material has been confirmed at thebottom of pressure vessel by muon
measurements. Possibility that fuelfallen in the lower plenum remains at
the bottom of RPV©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

34

2-12 Result of Muon Measurement

Unit 1Height OP (m)A shadow of the high density material that is considered to be fuel debris has been
confirmed at the bottom of pressure vessel.

Unit 2

Unit 335(Measurement result as of July 22, 2016)*Size of 1 pixel is equivalent to about 25 cm in reactor cross-section

North

Horizontal distance (m)SouthStructure of pressure vessel Lower Part

Considered that fuel fallen in the lower plenum remains at the bottom of RPV.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

Unit 1

Unit 2

Unit 32-13 Estimation of Pressure Vessel Lower Part・CRD on the outer periphery wasconfirmed during PCV internal
investigation, and the holes on
grating panel and some melting
material were observed whichindicate RPV bottom failure (notlarge, center and off-center
position)©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

36

Unit 1

2-14

Unit 2

37

Unit 3Result of Unit 2 Containment Internal
Investigation
In January and February 2017, investigations were conducted on containment vessels tocheck the status of debris falling on the pedestal platform and to control rod drive (CRD)
and the condition of pedestal structure.
Control rod drive (CRD) exchange rail

Pedestal opening

Pedestal
Control rod drive (CRD)

Primary containment vessel (PCV)

By guide Investigation area
pipe

Platform

Containment vessel through hole(×-6 penetration)
Underground level
Pedestal

X-6 pene
CRD rail

Pan tilt
cameraInvestigation area inside the pedestal

Collected much information on the situation in pedestal from theresults of processing images taken during the investigation with theguide pipe as shown on the left.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-15

Result of Unit 2 Containment Internal
Investigation

CRD housing
support bracket

CRD lower
part

Unit 1

Unit 2

38

Unit 3

TIP guide tube support

Grating dropped
PIP cable

CRD exchanger
CRD exchanger
Underground access

Slot opening

Platform
Pedestal
Flat barSlot
opening
CRD rail end(Reference) Inside the pedestal for Unit 5

Deformedgrating

Fallaout

Flat bar
TIP guide tube
support*Gap between CRD railand platform(about 150 to 40 mm)

TIP guide tube*(Reference) Photo of regular inspection inside thepedestal for Unit 2*For inspection of Unit 5, its TIP guide tube, and TIP
guide tube support have been removed

CRD rail end

Image of inside the pedestal looked through from CRD rail
Sharpening process allows us to view the position far away from camera
・Grating dropped on the left side of slot opening
・Steam rises from lower grating
・Grating not dropped out is observed to have adhered deposits and deformation

Note: Joints remain since the individually sharpened
images are patched (the same applies for the next
and subsequent pages)
TIP (Traversing In-core Probe) : Used for LPRM calibration
Images provided by International ResearchInstitute for Nuclear Decommissioning (IRID)
Image processed by TEPCO Holdings, Inc.

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-16

Result of Unit 2 Containment Internal
Investigation

CRD housing
support bracket

CRD lower
part

Unit 1

Unit 2

Unit 339

Area where PIP
and LPRM cables
is not locatable

PIP cable

CRD exchanger
Underground access

Slot opening

Platform
Pedestal
Flat bar

CRD rail end(Reference) Inside the pedestal for Unit 5

PIP cable
PIP guidetube support

PIP or LPRM cables+: Control rod drive(directly above PIP cables)□: SRNM (directly above SRNM cable)○: LPRM (directly above SRNM cable)□: TIP guide tube support
Projected the position of existing structure on grating•PIP or LPRM cables are observed on the far
side of the area where PIP or LPRM cables are
not locatable©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-17

Result of Unit 2 Containment Internal
Investigation

Unit 1

Unit 2

Unit 340

A clear boundary
cannot be found

CRD exchanger

Fallout like TIP
guide tube

Cable is confirmed on the far side
(not locatable)

Grating exists invisible area

Area where grating
status is unknown

Area where TIP guide tube canbe confirmed though PIP cablecannot be seen because no lightreaches (not locatable)

Cable-like fallout
Area where PIP or LPRM cables
are observed (not locatable)

Area where PIP or LPRM cables
cannot be identified

Area where PIP and LPRM cables
are locatable

Area where grating is
droppedArea where grating exists (remaining
grating is partially deformed)

Fallout

CRD rail
The area where grating dropped and the area where cables under the pressure vessel aredamaged generally match.
Deformed grating could possibly be associated with fuel debris fallen and damaged points otherthan those in the center of pressure vessel.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

2-18 Estimation of Pedestal Part

Unit 1

Unit 2

Unit 3•Observed steam rising during
the PCV internal investigation.

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

41

42

3. Unit 3 Estimation Diagram and its Basis©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

3-1

Decay Heat of Unit 3 and Heat
Generated by Water-Zr Reaction
Maximum value:

Unit 1

Unit 2

Unit 343

Decay heat
Heat generated by water-Zr reaction

Period without water injection

Water level drop
due to decay heatCoredamage/core
melt due towater-Zr reaction

RCIC/HPCI
cooling

Unit 3 became uncoolable because its HPCI stopped, which resulted in coredamage/meltdown during a no-injection period. Fire engine water injection
made no contribution to water-Zr reactions©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

3-2

Unit 3 Estimation Diagram

・Assuming the energy amount from
PCV pressure increased due tohydrogen generation, most of fuels
are estimated to have melted
・Since no temperature increase wasobserved in each part of RPV when
injection from CS system was stopped
from 9 to 24 in December 2013
(keeping constant total flow byincreasing flow from FDW), theamount of fuel debris in core isestimated to be small (smaller thanthat in Unit 2)
・Because temperature at RPV lower
head dropped when water injection
from CS system started (September 1,2011), fuel debris is estimated to inlower plenum
・Fuel rods can exist only on the outer
periphery if any
・Estimated to be general oxide debris
solidified with molten fuel･Fuel fallen in the lower plenum could
remain at the bottom of RPV
・A time delay before temperature risein response to reduction of water
injection volume indicates a possibilityof water retained in the pressure
vessel・Possibility that debris intrude in
CRD accompanying breakage of
CRGTUnit 1

Unit 2・Currently, the shroud has two
possibilities: sound or damaged.
・Possibility of pellets remaining on theouter periphery because the fueltemperature rise on the outer peripheral
may not be so highLegend44

Fuel rod
Oxide debris (porous)
Granular debris
Concrete mixed debris
Sound CRGT･If heat transfer from high temperaturemolten debris is small, CRGT will be
unmelted but remains

Damaged CRGT・Lower part of pressure vessel can bedamaged because MCCI is considered tooccur

CRD (debris inside)・Possibility of accumulation in stagnant
spot when there are particulate debris andpellets

・Possibility of part of fuel debris
solidified without causing MCCI

Unit 3・In addition to Unit 4, Unit 3 also
experienced an explosion to which hydrogengenerated by MCCI could possibly havecontributed.
・Meanwhile, DW spraying was performed formore than 1 h from 7:39 on March 13 inresponse to the accident. Diffusion of fueldebris could have been suppressedconsidering a water level on DW floor during
pressure vessel breakage.
・Estimated that fuel debris diffuses outside
the pedestal through the pedestal opening
but has not reached shell attack
・If water is pooled on PCV floor, particulate
debris can be formed.
・Possibility of accumulation in stagnant spot
when there are particulate debris

Sound CRD

Sound shroud
Pellets
Burning fuel*Oxide debris*
Heavy metal debris*
Powder pellets*
Clad residue*
Molten core internals*
Solidified B4C*
Control rod mixture melts**Not used in Illustration of Estimated Unit 3©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

3-3 Estimated State of Core

Unit 1

Unit 2

Unit 3•Assuming the energy
amount from the increased
PCV pressure by hydrogengeneration, it is estimatedthat most of fuels havemelted©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

45

Unit 1

3-4 Change in Reactor Water Level

Reactor water level (M)

HPCI auto start(3/12 12:35)

Unit 2

HPCI manualstop (3/13 2:42)

Unit 346

Fire engine water
injection started(3/13 9:25)?

Fuel area (A)Fuel area (B)Narrow range

Wide range

Wide range(corrected)

Fuel range(corrected)

・Before HPCI was manually stopped, the reactor pressure was lower than the HPCI operable range
and the HPCI discharge pressure was also low, which suggests a possibility that water injection into
the nuclear reactor could be insufficient before stopping HPCI・Fire engine water injection started at 9:25 on March 13, damage to fuel is considered to haveprogressed during the period when injection was not allowed.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

3-5

Unit 1

Behavior of Reactor
Pressure/Containment Vessel Pressure

Unit 2

47

Unit 3

2.2
D/W
S/C
D/W (operator data)
S/C (operator data)
RPV
RPV (operator data)

Reactor pressure/containment
vessel pressure (MPa[abs])
原子炉圧力・格納容器圧力

2.0
1.8
1.6
1.4
1.2
1.0
0.8
0.6
0.4
0.2
0.0
3/13
6:00

3/13
7:00

3/13
8:00

3/13
9:00

3/13
10:00

3/13
11:00

3/13
12:00

3/13
13:00

3/13
14:00

3/13
15:00

At around 9 o’clock and past 12 o’clock on March 13, 2011, the containment pressure
has increased significantly.
Like Unit 2, this could be caused by a large amount of hydrogen generated there andmost of the fuel could have melted.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

3-6 Estimation of Pedestal Part

Unit 1

Unit 2

Unit 3・In addition to Unit 4, Unit 3 also
experienced an explosion to whichhydrogen generated by MCCI could
possibly have contributed.
・Meanwhile, DW spraying was
performed for more than 1 h from
7:39 on March 13 in response to theaccident. Diffusion of fuel debris
could have been suppressedconsidering a water level on DW
floor during pressure vessel
breakage.
・Estimated that fuel debris diffuses
outside the pedestal through thepedestal opening but has notreached shell attack.

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

48

Unit 1

3-7 Flow into Unit 4 from Unit 3

About 0.1 mSv/h

About 6.7 mSv/h

From reactor building

Unit 3 emergency gas
Processing system exhaust pipe

About 0.8 mSv/h

Emergency gas treatment system
Ventilator (A)From pressure control room

Inlet side

Unit 349

Exhaust stack

About 0.5 mSv/h

Possibility of reverse flow

Unit 2

Emergency gas treatment systemradioactivity removal filter

To exhaust stack

Outlet side
Emergency gas treatment system
Ventilator (B)SGTS filter train
About 0.5 mSv/h
A high dose on the outlet side
About 0.1 mSv/h
About 5.5 mSv/h

In Unit 4, hydrogen explosion occurred in thereactor building around 6:14 on March 15.
The vent gas of Unit 3 is considered to begetting into the Unit 4 reactor building.
Units 3 and 4 exploded due to hydrogengenerated in Unit 3. In addition to thehydrogen generated by water-Zr reaction, thehydrogen generated by MCCI could possibly
have contributed to explosion.

©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

3-8 Investigation Result of MSIV Room

Unit 1

Unit 3

Unit 2

50

Leakage is observed from the periphery of the expansion joint of main steam pipe D.
Leaked water is estimated to be the volume of 2 to 4 pencils.
Main steam pipe C
Expansion joint C
Expansion joint C

Expansion joint D
Leakage

Main steam pipe D (far side of Main steam pipe C)
Enlarged view of leakage
MS process monitor

HPCI feed water
pipe
Main steam piping

Grating

Expansion joint
Grating

PCV side

Feed water piping

Spare (×-46, 47)
Feed water system pipe penetration(×-9A,B)

Main steam pipe D

Expansion joint

Main steam system pipe penetration (X-7A～D)

Leakage

Main steam system drain pipe penetration (X-8)

View of MSIV room from PCV side

Image of MSIV room (cross-section)

At Unit 3, water level in the containment vessel is relatively high (about 6 m) compared to Units 1
and 2. May 15, 2014, after the investigation of MSIV room, leakage of water is observed from theperiphery of the expansion joint of main steam pipe D.
→ Because the containment vessel maintains a high water level, it is not considered to have reachedshell attack.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

４ Summary and Future Plan

51

[Summary]
TEPCO has been making the best effort to estimate the fuel debris
distribution in Unit 1 to 3 based on the various information and the latest
knowledge (for example：the data obtained during the accident, the dataobtained during the cooling of fuel debris after the accident, the results ofon-site investigation, and the results of the accident progression
analysis).
[Future Plan]TEPCO will continue to collect new information acquired in on-site
investigation etc. and reflect it appropriately to the estimation.
In some cases, it may be necessary to change the estimations in thispresentation. But TEPCO will improve the understanding of the status
inside RPV and PCV at Fukushima Daiichi NPS through these efforts andwill use it to determine the policies and methods for the fuel debris
removal.
©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.

52

Reference：Glossary
BWR：Boiling Water Reactor
CRD：Control Rod Drive
CRGT：Control Rod Guide Tube
CS：Core Spray
D/W：Drywell
FDW：Feed Water
GOTHIC：Generation of Thermal Hydraulic Information in Containments
HPCI：High Pressure Coolant Injection System
HVH：House Ventilation and Heating System
IC：Isolation Condenser
LPRM：Local Power Range Monitor
MAAP：Modular Accident Analysis Program
MCCI：Molten Core Concrete Interaction
MSIV：Main Steam Isolation Valve
PCV：Primary Containment Vessel
PIP：Position Indicator Probe
PLR：Primary Loop Recirculation System
RPV：Reactor Pressure Vessel
RCIC：Reactor Core Isolation Cooling System
RCW：Reactor Building Closed Cooling Water System
S/C：Suppression Chamber
SRV：Safety Relief Valve
TIP：Traversing Incore Prove System
Zr：Zirconium©Tokyo Electric Power Company Holdings, Inc. All Rights Reserved.