This document is a translated version of the original document in Japanesepublished by the Tokyo Electric Power Company on its web site below. The
translation is done by the Institute of Applied Energy. The original version
in Japanese is the official document.
http://www.tepco.co.jp/nu/fukushima-np/images/handouts_111130_09-j.pdf

Core conditions of Unit-1 to Unit-3 of the
Fukushima Daiichi Nuclear Power Station

November 30, 2011
Tokyo Electric Power Company

Summary

Estimated core conditions of Unit-1 to Unit-3 of the Fukushima Daiichi Nuclear Power
Station were disclosed on May 23, 2011, which explained that the cores of all units had been
largely damaged, the molten fuel had been relocated to the lower plenum, and most of it had
been cooled in the vicinity of the lower plenum, although a possibility of some molten fuelhaving been relocated to outside the reactor pressure vessel (RPV) could not be excluded.
The disclosed estimation came from a comprehensive judgement based on the core
conditions predicted by the MAAP analyses and the core conditions derived from themeasured temperature behaviors at various locations in the containment vessels (PCVs)(see Figure 3.1-1).

Since the above estimation was made in May 2011, a number of new operational actions(see ① below, for instance), examinations and analyses have been attempted forinvestigation. These results have provided updated findings on estimations of the core
conditions. These updated estimations include the following.
①

It could be estimated from temperature behavior observed at various points when
the water injection paths or amounts were changed, that: in Unit-1, little fuel debris
was left in the RPV, because the RPV temperatures decreased largely; and at Unit-2
and Unit-3, some of the fuel debris was present in the RPV.

②

It could be estimated from the results of water filling to the variable and reference
legs of the reactor water level indicators and calibration of the water level indicators,
that: in Unit-1 and Unit-2, the reactor water level was not in the original core region
and the fuel was not present at its original position.

③

It could be estimated from Cs concentrations obtained in the nuclei analysis ofgases in the PCVs of Unit-1 and Unit-2, that: in Unit-1 the amount of molten fuel waslarger than that of Unit-2.

④

It could be estimated from the heat balance evaluation of decay heat and heatremoval, i.e., the initial decay heat which could not be removed by the isolation
condensers (ICs) or high pressure core injection (HPCI) systems, that: in Unit-1 coredamage and RPV damage occurred earlier than in Unit-2 or Unit-3, and the initial
decay heat of Unit-1 that could not be removed was about three times that of Unit-2
or Unit-3.

⑤

It could be estimated from the heat balance evaluation in the RPV, that: in Unit-2
and Unit-3, the fuel was mostly under water and the uncovered fuel was less than
3% as of October 10, 2011.

⑥

It could be estimated from the analysis of core-concrete reactions, that: even in
Unit-1, which was considered to have the largest fraction of fuel being relocated to
the PCV pedestal, the erosion depths of PCV pedestal concrete were not so deep
as to be reaching the PCV inner wall.

Comprehensive analysis of the information thus obtained could update the estimated core
conditions of May 2011. In Unit-1, almost all molten fuel in the accident progression was
relocated to the reactor vessel lower plenum, leaving almost no fuel in the original coreregion. Most of the fuel debris in the lower plenum was considered to have been further
relocated to the PCV pedestal, where the fuel debris caused core-concrete reactions, but asof November 2011 (the time of this report compilation) the core-concrete reactions were
estimated to have ceased because the debris has been cooled by injected water and its
decay heat has decreased. The debris was considered now (November 2011) to be coolingstably. In Unit-2 and Unit-3, the molten fuel was estimated to remain partly in the originalcore region, to have been relocated partly to the reactor lower plenum, and now (November2011) to be cooling stably.
It should be noted that the conditions inside the RPV or the PCV have not been directly
and visually observed, but indirectly estimated based on indirectly acquired information.
Possibilities of direct visual observation will be pursued hereafter.

Table of contents

1. Introduction ................................................................................................................ 1
2. Findings from analyses ................................................................................................. 2
2.1 MAAP analysis ........................................................................................................... 2
2.2 Heat balance between decay heat and heat removal................................................. 3
3. Findings from observed results..................................................................................... 3
3.1 Estimation from measured temperatures and pressures ............................................ 3
3.2 Heat balance in the RPV ............................................................................................ 4
3.3 Readings of reactor water level indicators .................................................................. 5
3.4 Nuclei analysis of in-PCV gases ................................................................................ 5
3.5 Findings from other observed results ......................................................................... 6
4. Impacts of core-concrete reactions on the containment vessels .................................. 7
4.1 Core-concrete reactions ............................................................................................. 7
4.2 Reactor building closed cooling water system (RCW) of Unit-1 ............................ 8
4.3 Evaluation results of core-concrete reactions ............................................................. 8
4.4 PCV gas analyses ...................................................................................................... 9
5. Cooling situations of each unit ...................................................................................... 9
5.1 Cooling situations of Unit-1 ........................................................................................ 9
5.2 Cooling situations of Unit-2 ...................................................................................... 10
5.3 Cooling situations of Unit-3 ...................................................................................... 11
6. Estimation of core conditions ...................................................................................... 11
6.1 Core conditions of Unit-1 .......................................................................................... 11
6.2 Core conditions of Unit-2 .......................................................................................... 12
6.3 Core conditions of Unit-3 .......................................................................................... 12

1. Introduction
The Tohoku District Off-Pacific Ocean Earthquake (Great East Japan Earthquake) which
occurred on March 11, 2011 with its hypocenter off the Sanriku Coast, led to a series of
events caused by the earthquake (tsunami, station blackout coupled with damage toemergency power generation equipment and subsequent loss of cooling equipment) in
Unit-1 to Unit-3 of the Fukushima Daiichi Nuclear Power Station that caused a severe
accident as the units remained in continued station blackout conditions. The accident far
exceeded design basis accidents and even multiple accidents which had been assumed indeveloping the accident management procedures, i.e., all emergency core cooling systems
could not function or stopped including those of adjacent units. It is very important to grasp
both the accident progression after the earthquake and the current plant conditions fortermination of the accident and for recovery activities hereafter.
In response to a request from the Nuclear and Industrial Safety Agency (NISA), theMinistry of Economy, Trade and Industry (METI) (*) received on April 25, 2011, plant data at
the time of the earthquake were collected to the extent possible and were reported to METI(**) on May 16, 2011. The plant statuses were evaluated by the Modular Accident Analysis
Program (hereafter called “MAAP”) by using the collected data. The results were submitted
on May 23, 2011, as the Annex (***) to the data report above.
Work has been continued for recovery at Unit-1 to Unit-3 since the report was submitted.
Consequently, temperatures and pressures in the RPV and PCV have decreased and stablecooling has become possible. During the 6 months after the report submission, new findings
have been accumulated concerning the reactor behavior through the experienced changes
of water injection methods to the reactor, changes of injection water amounts and changesof environmental conditions including natural phenomena. Among such new findings, someare not consistent with the past estimation of reactor conditions. Therefore, the accumulated
findings were reviewed and the estimation of “Core conditions at Unit-1 to Unit-3 of the
Fukushima Daiichi Nuclear Power Station” has been updated.
It should be noted that the results reported here were based on the limited information
available at the time of this report compilation and some estimates and assumptions weremade as needed for analyses. The uncertainties in the analysis results may be significant.
The results therefore can be quite different from the results which will be obtained hereafter
based on future investigations and evaluations.
(*)

Request of reports pursuant to Clause 1, Article 67 of the Reactor Regulation Law(Law No. 67) (Notification Number: H23-04-24 Gen-1), April 25, 2011, NISA, METI

1

(**)

Operation records, analyses of accident records and evaluation of their impacts at
the Fukushima Daiichi Nuclear Power Station at the time of the Tohoku District
Off-Pacific Ocean Earthquake, May 16, 2011, Tokyo Electric Power Company(***)

Estimated core conditions at Unit-1 to Unit-3 of the Fukushima Daiichi Nuclear
Power Station, May 23, 2011, Tokyo Electric Power Company）

2 Findings from analyses
2.1 MAAP analysis
The MAAP analysis gave the following results. In Unit-1 the core was damaged at a fairly
early stage leading to the RPV damage thereafter, if the isolation condensers (ICs) wereassumed to have stopped after the station blackout due to the tsunami arrival. In addition, it
has been found from the calibration of Unit-1 reactor water level indicators that the waterlevel in the RPV was not in the core region, notwithstanding the readings of the water levelindicators. On the other hand, in Unit-2 and Unit-3 the cores were damaged but eventuallyretained in the RPVs; the cores were damaged when the reactor water levels decreased
after the reactor core isolation cooling (RCIC) systems or high pressure coolant injection(HPCI) systems had stopped, although reactor water injection could have been continued
after the station blackout until the tsunami arrival; but the cores could have been eventuallyretained in the core region owing to the restarted reactor water injection. But it should be
noted that the readings of water level indicators might be incorrect, because the water in thewater level indicators might have evaporated. Assuming the lower actual reactor water level
was lower than the indicated readings, cutting below the bottom of active fuel (BAF), MAAP
predicted further advancing of the core damage, leading to the RPV damage thereafter.
Figure 2.1-1 illustrates the fuel distributions in the core of each unit at the time theanalysis was terminated.
It was estimated that the fuel was being cooled after being relocated from the core region
to the lower plenum, because the following data were measured in Unit-1 to Unit-3 at thetime of this report on MAAP results (May 2011): temperatures in the RPV bottom, thesewould not have been measurable if the RPVs were significantly damaged; hightemperatures, these would be likely when a heat source existed in the RPVs; and multiple
temperature indicators had value changes at their measurement points consistent with thechanges in the amounts of reactor water injection.
(Refer to Attachment-1 “Estimated core conditions disclosed in May 2011”)

2

2.2 Heat balance between decay heat and heat removal
Unit-1 to Unit-3 were in a situation that they were unable to remove their decay heat until
water injection to the reactor could be started after heat removal by the ICs, RCIC or HPCI
had stopped. Consequently, the fuel was overheated, and the core was damaged in each
unit. The decay heat decreases rapidly as soon as the fission reactions cease due to thereactor scrams. Figure 2.2-1 compares decay heat not removed between the three units In
Unit-1, the ICs stopped early, and more time was needed to start water injection to thereactor. The decay heat that had not been able to be removed was about three times that of
Unit-2 or Unit-3. Figure 2.2-2 indicates the decay heat not removed from Unit-1 was bigger
than the energy needed for melting fuel or structures, while it was less at Unit-2 and Unit-3.
This was the main reason for the present MAAP results (Case 1): the core damage started
earlier at Unit-1, and the RPV was damaged; on the other hand, early initiation of water
injection could have retained fuel in the core region of Unit-2 or Unit-3. At Unit-2 and Unit-3,
a sufficient amount of water injection was assumed to have been secured once injection had
been started. If this assumption failed and the decay heat could not be adequately removed,
the core damage would advance and the RPV would be damaged as predicted by MAAP
(Case 2).
(Refer to Attachment-2 “Core conditions estimated from heat balance while water injection
was being interrupted”)

3. Findings from observed results
3.1 Estimation from measured temperatures and pressures
Figure 3.1-1 shows temperature changes at typical points of Unit-1. Water was injected
via the feedwater line which did not go through the core directly, but the measured
temperatures cut below 100 deg C as of August 2011. It was considered, therefore, that thefuel was being sufficiently cooled in the RPV lower plenum or the PCV pedestal.
Figures 3.1-2 and 3.1-3 show temperature changes at typical points of Unit-2 and Unit-3,
respectively. For Unit-2 and Unit-3, the temperature changes after the accident indicatedthat the temperatures remained higher at the RPV upper part than at the RPV lower part.
Since the RPV water level was considered to have been below the core level, part of the fuelwas considered to have been present in the gaseous phase in the core region, i.e., steam
generated from the injected water in the RPV lower part was considered to have beenheated up by the uncovered core fuel and that raised the temperatures at the RPV upper
part. The uncovered fuel was considered to have remained in the core region, because the3fuel in the outer core region had not melted due to lower power densities and therefore lower
decay heat. Even if the cladding of uncovered fuel had kept its configuration when the water
injection was initiated, its original configuration was likely to have been lost at the time of thisreport (November 2011), because it (the uncovered fuel) remained overheated and exposedto the steam environment over an extended time. Water injection from core spray (CS) lines
immediately above the core region started on September 1 in Unit-2 and September 14 in
Unit-3. Consequently, cooling of the uncovered fuel left in the core region could have started
and temperatures at all measurement points could have significantly dropped. Thisestimation was consistent with the MAAP analysis results that the whole fuel of Unit-1 was
relocated from the core region, while part of the fuel remained in the outer core region in
Unit-2 and Unit-3.
(Refer to Attachment-3 “Core conditions estimated from the measured temperatures andpressures”)

3.2 Heat balance in the RPV
The water injected into the reactor was discharged as liquid or steam after being heatedby decay heat. By using this scenario, the decay heat energy consumption in the RPV (heatbalance) was schematized in a model as indicated in Figure 3.2-1. Based on this model,
possible core conditions were investigated which could reproduce measured temperature
increases. Five forms of energy consumption were considered: ① water temperature
increases; ② water evaporation; ③steam temperature increases; ④ fuel temperature
increases; and ⑤ temperature increases of structural materials. For a given amount ofwater injection to the reactor and the decay heat, reactor conditions could be estimated, byusing the assumed heat balance model, which can reproduce the measured parameters.
The investigation predicted the fraction of uncovered fuel as of October 10, 2011, was about
3% or less for Unit-2 and Unit-3, the fuel being mostly covered by water. It should be notedthat this type of investigation was not conducted for Unit-1, because the current heatbalance model assumed energy transfer from fuel to structural materials by steam produced,
and the situation of a small amount of steam production was not covered, while thetemperatures around the RPV of Unit-1 were low (and steam production was limited).
(Refer to Attachment-4 “Core fuel temperatures estimated by the in-RPV temperatureevaluation model”)

4

3.3 Readings of reactor water level indicators
Figure 3.3-1 shows the reactor water level indicator configuration. As seen in the figure,the water level in the reference leg located outside the RPV is kept constant, and thepressure difference between the water head in the reference leg and the water headcorresponding to the reactor water level (Hs – Hr) is converted to indicate the reactor waterlevel. However, under accident conditions, the water in the instrumentation lines may
evaporate. If the water in the reference leg evaporates, for instance, the reference waterlevel lowers and the reactor water level will read higher (Figure 3.3-2).
On May 11 for Unit-1, reactor water level indicators were calibrated by installing atemporary differential pressure gauge, and filling water to the reference leg andinstrumentation line. The reactor water level indicators showed that the reactor water level
was below the top of active fuel (TAF), at minus 5 m or lower. For Unit-2, a differential
pressure gauge was installed on June 22, and the reference leg and the variable leg werefilled with water on June 22 and again on October 21. For Unit-2, calibration of the waterlevel indicators could not be conducted because of the high radiation level, but the reactor
water level was estimated to be 5 m below TAF-5m or lower from instantaneous readings oftemporary differential pressure gauges installed after the accident. It should be noted thatthe water in both lines of the reference leg and the variable leg had been observed toevaporate within a short time after filling on June 22, and that the water in the variable leg
had been observed to evaporate gradually after it was filled on October 21.
From these observations, the reactor water levels are considered (as of this reportcompilation) not yet to have reached the original fuel region in both Unit-1 and Unit-2, andthe fuel was considered unlikely to have remained at its original position of the original
configuration. In Unit-2, part of the fuel (acting as a heat source) might be present in thevicinity of the variable leg, because only the water in the variable leg evaporated in Octoberwhen water was filled. In Unit-1, water evaporation was not observed in the variable leg.
In Unit-3, water level indicator calibration and water filling have not been conducted due toextremely high-level radiation near the instrumentation devices.
(Refer to Attachment-5 “Calibration of water level indicators”)

3.4 Nuclei analysis of in-PCV gases
Gamma rays from nuclei in the gases in the Unit-1 PCV and the Unit-2 PCV wereanalyzed. Table 3.4-1 shows the results. As summarized in the table, the Cs radioactivity
concentration (corrected) in the Unit-1 PCV was about three times that in the Unit-2 PCV.
The Cs discharge rate might depend on the gas fractions and temperatures in the PCV.

5

Therefore, direct comparisons were not possible, but Unit-1 shows the severest results. Thiswas consistent with the results from other evaluations that the core damage of Unit-1 was
the largest.
No gas sampling was conducted for Unit-3, because the radiation level was too high near
the sampling lines.
(Refer to Attachment-6 “Radioactivity concentrations in the atmosphere of thecontainment vessel”)

3.5 Findings from other observed results
The following findings have been observed besides those in 3.1 to 3.4. Some of them aredifficult to use for estimating the core conditions, and some others may be effectively used,
but no conclusive knowledge has been derived yet. Analysis work and estimation will be
continued.
①

Condition checking of LPRM detectors (Unit-2, Unit-3)
TDR measurements (time-domain reflectometry: a measurement technique used todiagnose disconnection/insulation deterioration of electrical lines by observing reflected
waveforms) were conducted on the LPRM detectors, which were one of the neutron
monitors installed in the reactors. Attempts were made to estimate the reactor bottom
conditions from the measurement results, but at the time of this report compilation, it
has been found not to be easy to derive convincing clues.

②

Condition checking of control rod position indicator probes (PIPs) (Unit-1, Unit-3)
Energization checks were conducted on the control rod position indicator probes(PIPs), which were in-core position monitors of control rods mounted on the control rod
drive mechanism. Attempts were made to estimate the reactor bottom conditions fromthe measurement results, but at the time of this report compilation, it has been found
not to be easy to derive convincing clues.

③

Recovery work of drywell (D/W) equipment sump thermometers (Unit-1, Unit-2,
Unit-3)In-service measurements by D/W equipment sump thermometers were attempted inorder to get the PCV bottom temperatures. Temperatures were measured for Unit-1 andUnit-3, but the instrumentation line was found to be disconnected for Unit-2. Continued
analysis is considered to be necessary, because these thermometers have not been inservice for a long period and no definite trends have been observed yet.

④

Recovery work of the primary loop recirculation (PLR) pump inlet thermometers(Unit-1, Unit-2, Unit-3)

6

In-service measurements by the primary loop recirculation (PLR) pump inlet
thermometers were attempted in order to get the PCV bottom temperatures.
Temperatures were measured for all units. Continued analysis is considered to benecessary, because reliabilities of readings of these thermometers were still being
analyzed at the time of this report compilation.
(Refer to Attachment-7 “Operability checks of local power range monitor (LPRM)
detectors (Unit-2, Unit-3) (Attempt to estimate core conditions from LPRM data)”)(Refer to Attachment-8 “Operability checks of control rod position indicator probes (PIPs)(Unit-1, Unit-3) (Attempt to estimate core conditions from control rod position indicator
probe (PIP) data)”)(Refer to Attachment-9 “Results of condition checks and behavior of drywell (D/W) sumpthermometers”)(Refer to Attachment-10 “Behavior of the primary loop recirculation (PLR) pump inlet
thermometers”)

4. Impacts of core-concrete reactions on the containment vessels
4.1 Core-concrete reactions
Upon relocation of molten fuel to the PCV, it will spread on the pedestal floor if the fluidity
is maintained, part of the molten fuel further will leak out through pedestal slits and solidify
as flat lumps with a large surface (Figure 4.1-1). If there are openings on the pedestal floor,
such as an equipment drain sump pit, the fuel debris may clog them heavily (Figure 4.1-2). If
water is retained on the bottom of the PCV, the molten fuel will solidify, being cooled upon
contact with water, into many small lumps. Thus, there are large uncertainties in shapes anddistributions of fuel debris, once the molten fuel drops to the PCV. Large uncertainties also
exist in heat transfer from the fuel debris to water, since the fuel debris can contact water inmany diverse configurations. If the heat of fuel debris in the PCV cannot be sufficiently
removed, core-concrete reactions occur, in which the concrete is heated to above its melting
temperature, and the concrete is thermally decomposed and eroded. The erosion ends
when the decay heat decreases and water injection to the reactor is resumed. The erosion
depths vary greatly depending upon conditions assumed, for instance, uncertain
geometrical configurations (an easily heat-removable flat shape or a hard to remove pit-type
shape) or heat transfers.

74.2 Reactor building closed cooling water system (RCW) of Unit-1
Radiation distributions measured in the Unit-1 reactor building (R/B) showed high dose
rates on the RCW system lines (Figure 4.2-1). The RCW is a closed loop for coolingauxiliary equipment and it is unlikely to be contaminated as high as several hundreds of
mSv/h in normal situations. But the RCW lines were laid widely in the R/B and they cooled
the equipment in the PCV, too. As seen in Figure 4.2-2, the RCW line for drain cooling waslaid in the equipment drain pit in the lower part of the PCV. Therefore, it was highly possible
in Unit-1 that the molten fuel was relocated to the equipment drain pit and damaged the
RCW piping and this caused the high RCW line contamination. It was considered that, upondamage to piping, high dose steam or water was transferred to the RCW secondary lines
and radioactive materials were simultaneously transferred. On the other hand, if the RCWlines had been damaged by the relocated fuel debris, water from the RCW secondary lines
might have flowed to the PCV and contributed to the fuel debris cooling.
No such high dose rates have been recognized for Unit-2 and Unit-3.
(Refer to Attachment-11 “Contamination of the reactor building closed cooling water(RCW) system”)

4.3 Evaluation results of core-concrete reactions
MAAP has a module which can evaluate core-concrete reactions, too, not only the module
to evaluate fuel behavior in the RPV. This module for core-concrete reactions was used toanalyze the Unit-1 core-concrete reactions. Unit-1 was thought to have the largest fraction of
relocated fuel debris to the PCV. Big uncertainties in the initial conditions and given
analytical conditions might cause big uncertainties in the results as well. Realistic conditions
shown in Table 4.3-1 were used in the evaluation and the results are shown in Figure 4.3-1.
The results indicated that the concrete was eroded but the fuel debris could remain in the
PCV.
A seismic evaluation was also conducted when the pedestal floor concrete had beeneroded by core-concrete reactions. No seismic concerns were identified.
(Refer to Attachment-12 “Impacts of core-concrete reactions on the reactor containment
vessel”)(Refer to Attachment-13 “Estimation of the conditions of structural materials in the Unit-1
containment vessel”)

8

4.4 PCV gas analyses
Gas samples used for nuclei analysis in 3.4 were also provided for gas composition
analysis. It was long after the core-concrete reactions were thought to have ceased, andputting aside the question of whether or not they had occurred, that gas sampling from thePCV became possible. Therefore, even if hydrogen, carbon monoxide or carbon dioxide had
been generated in the core-concrete reactions, they would be unlikely to have remained in
the sampled gas at the time of gas sampling, since they were diluted with steam, nitrogengas and others. Table 4.4-1 shows the results of gas analysis. In all samples, theconcentration of carbon dioxide was limited to the level which would be anticipated whencarbon dioxide contained in the reactor water was separated and transferred to the gaseous
phase, while the concentrations of hydrogen and carbon monoxide were at the level of thedetection limit. This means that no core-concrete reactions were occurring at the time of thisreport compilation.
(Refer to Attachment-14 “Results of in-containment gas composition analysis”)(Refer to Attachment-15 “Gas residues from the early phase of core-concrete reactions”)

5. Cooling situations of each unit
5.1 Cooling situations of Unit-1
The RPV temperatures of Unit-1 were lowered to about 40 deg C as of November 21,
2011, as seen in Figure 5.1-1. No symptoms of superheated steam production wererecognizable in the core region where injected water could not reach.
Steam blowout from the Floor 1 penetration observed on June 3, 2011, was notrecognizable on October 13, 2011 (Figure 5.1-2 and Figure 5.1-3). It was concluded thatcooling was being continued with decreasing decay heat. Another point to be noted was that
the RPV temperatures and D/W temperatures decreased, and that the suppression
chamber (S/C) pool temperatures slightly increased when the amount of water injection wasincreased from October 28, 2011 (Figure 5.1-4). This might indicate the following scenario:
steam was being produced in the PCV but it was condensed before leaking out to the R/B
until the injected water amount was increased; and more decay heat was consumed forincreasing water temperature, steam production decreased, more hot water flowed into the
S/C, and the S/C temperature increased. The targeted increase of injected water amount
was the flow rate that could remove decay heat only by water temperature increases without
steam production. The temperature behavior anticipated as its consequence was actuallyobserved, and therefore, well-controlled cooling was being implemented.

9Figure 5.1-5 shows the changes of D/W pressure and nitrogen gas filling pressure, whichwere recorded for monitoring the nitrogen gas injection conditions. The nitrogen gas fillingpressure should behave as the D/W pressure if the injection nozzle is in the gaseous phase,because the pressures in the D/W should be uniform, but if the injection nozzle is covered
with water, the filling pressure should be higher than the D/W pressure, because the fillingpressure must be higher than the D/W gaseous pressure plus the water head. Figure 5.1-5shows that, after the amount of injected water to Unit-1 was increased on October 28, 2011,
the nitrogen gas filing pressure and the D/W pressure started to move away from each otherfrom around November 1, 2011. Thereafter, both pressures reached the same level, despite
no changes in the amount of injected water. The elevation of the nitrogen gas filling nozzle
was OP6700mm, while the bottom end of the vent tube, which is the theoretical minimum
water level from structural constraints, was OP6600mm. Therefore, the current PCV waterlevel was considered to be between these two levels, but there was no way to determine thedefinite water level at the time of report compilation.

5.2 Cooling situations of Unit-2
The Unit-2 RPV temperatures were lowered to about 80 deg C as of November 21, 2011,as shown in Figure 5.2-1. This temperature decrease was considered to be due to the water
injection via the CS system from September 14, 2011. The water from the CS could cool thefuel debris left in the core region, which was in the water injection path of the CS.
Steam blowout from the R/B Floor 5 just above the reactor that was observed on
September 17, 2011 was not recognizable on October 20, 2011 (Figure 5.2-2 and Figure
5.2-3). It could be concluded that cooling was being continued owing to the increased water
injection from October 4, 2011. In the photo of October 20, 2011, significant deterioration of
coatings on the ceiling crane was recognized. The coatings were considered to have lostadhesiveness by absorbing moisture and they were stripped off due to increased internal
stresses caused by dehydration. This observation also indicated that steam release fromimmediately above the reactor had ceased.
It was attempted to estimate the PCV water level, as was done for Unit-1, by comparing
the D/W pressures with the pressures at the nitrogen gas discharge nozzle, which wereconsidered to include the water head above the nozzle. It was not possible, though,because no appropriate discharge nozzle was located. The fuel in the PCV was judged
probably to be under water, because the amount of relocated fuel of Unit-2 was consideredlimited, sufficient water was now being injected for removing decay heat only by sensible

10

heat, and no unusual hot spots were observed in the PCV atmospheric temperature
measurements.
(Refer to Attachment-16 “Paint stripping-off incidents of Unit-2 reactor building ceiling
cranes”)

5.3 Cooling situations of Unit-3
The Unit-3 RPV temperatures were lowered to about 70 deg C as of November 21, 2011,as shown in Figure 5.3-1. This temperature decrease was considered to be due to the water
injection via the CS system from September 1, 2011. The water from the CS could cool thefuel debris left in the core region, which was in the water injection path of the CS.
The temperature increase observed around the shield plug probably was caused by thesteam blowout on March 20, 2011. As of October 20, 2011, the number of high temperature
spots had decreased, and the scale became smaller (Figure 5.3-2 and Figure 5.3-3). It could
be concluded that cooling was being continued with decreasing decay heat.
Figure 5.3-4 compares the D/W pressure and S/C pressure. If the S/C nitrogen gas
discharge nozzle is not under water, the S/C pressure should behave as the D/W pressure,but if the discharge nozzle is covered by water, the S/C pressure should exceed the D/Wpressure in order to withstand the D/W pressure plus the water head. Figure 5.3-4 showsthat the S/C pressure was constantly higher than the D/W pressure from October 1, 2011.
The pressure difference between the two indicated that the current PCV water level was at alevel between OP12,000 and OP13,000. The fuel in the PCV would probably be under water,
because the amount of relocated fuel of Unit-3 was considered limited, sufficient water was
now being injected for removing decay heat only by sensible heat, and no unusual hot spots
were observed in the PCV atmospheric temperature measurements.

6. Estimation of core conditions
6.1 Core conditions of Unit-1
At Unit-1, water was being injected via the feedwater system (FWS), as seen in Figure6.1-1, and the water injected to the RPV was flowing down outside the shroud and reaching
the lower plenum. The reactor water level indicator calibration (3.3) showed that the RPV
water level was below TAF-5m, namely, the water level was not in the core region.
Based on such findings and the evaluation results in 5.1 and other places, the Unit-1 core
conditions could be estimated as shown in Figure 6.1-1: almost all the molten fuel in theaccident progression was relocated to the RPV lower plenum, leaving almost no fuel in theoriginal core region. Most of the fuel debris in the lower plenum was considered to have11

been further relocated to the PCV pedestal. The fuel debris caused core-concrete reactions,but by November 2011, the time of this report compilation, the core-concrete reactions were
estimated to have ceased because the debris was being cooled by injected water and its
decay heat decreased, and the debris was considered to remain in the PCV.

6.2 Core conditions of Unit-2
At Unit-2, water was being injected via the CS and the FWS, as seen in Figure 6.2-1. The
water injected from the CS to the RPV flowed down inside the shroud, while the water fromthe FWS flowed down outside the shroud, and both reached the lower plenum. The resultsof water filling to the water level indicators (described in 3.3) indicated the RPV water level
was below TAF-5m, and no water level was considered to be present in the core region.
Based on such findings and the evaluation results in 5.2 and other places, the Unit-2 core
conditions could be estimated as shown in Figure 6.2-1: part of the molten fuel in theaccident progression was relocated to the RPV lower plenum or PCV pedestal, leaving partof the fuel in the original core region.

6.3 Core conditions of Unit-3
At Unit-3, water was being injected via the CS and the FWS, as seen in Figure 6.3-1. The
water injected from the CS to the RPV flowed down inside the shroud, while the water fromthe FWS flowed down outside the shroud, and both reached the lower plenum. The Unit-3
RPV temperatures were lowered to about 70 deg C as of November 21, 2011. Thistemperature decrease was considered to be due to the water injection via the CS systemfrom September 1, 2011. The water from the CS could cool the fuel debris left in the coreregion, which was in the water injection path of the CS.
Based on such findings and the evaluation results of 5.3, the Unit-3 core conditions could
be estimated as shown in Figure 6.3-1: part of the molten fuel in the accident progression
was relocated to the RPV lower plenum or PCV pedestal, leaving part of the fuel in theoriginal core region.

End

12

Table 3.4.1

Estimated concentrations in PCV gases
Radioactivity concentrations (Bq/cm3)

Nuclei

Unit-1

Unit-2(Sampled on 9/14/2011)(Sampled on 8/9/2011)

Cs-134

1.6×100

4.4×10-1

Cs-1372.0×100

4.6×10-1

Steam fraction

About 46%

About 100%

Table 4.3-1

Conditions for core-concrete reaction analysis
Conditions

Initial amount of fuel debris

Unit-1: Full core (100%)

Decay heat source

ORIGEN2 data with fuel burnup history being
considered

Depletion of volatile FP decay

20% depletion assumed

heat
Fine grain formation of fuelNot considereddebris by water originallypresent in the pedestal
Fuel debris sedimentation

Inflow conditions to sump pits:
P/D uniform sedimentationon the floor
D/W partial outflow
from the floor

Fuel debris into sump pits
Fuel debris spreading
on P/D, D/W floors
Fuel debris sedimentation

In sumps: 0.81m

thickness(P/D, D/W floors: 0.35m）

13

Table 4.4-1

Concentrations (%) of hydrogen, carbon monoxide, carbon dioxide in PCVs

H

CO

CO2

Unit-2 (August 2011) Sample ①

0.558

0.016

0.152

Unit-2 (August 2011) Sample ②

1.062

0.017

0.150

Unit-2 (August 2011) Sample ③<0.001

<0.01

0.152

Unit-1 (September 2011) Sample ①

0.154

<0.01

0.118

Unit-1 (September 2011) Sample ②

0.101

<0.01

0.201

Unit-1 (September 2011) Sample ③

0.079

<0.01

0.129

14

Core damage schematic↑: No fuel (collapsed): Intact fuel

Fuel zone燃: Sedimented damaged fuel (pin shape being kept): Fuel pin diameters enlarged by solidified molten
fuel while going down pin surfaces: Coolant flow paths blocked by enlarged fuel pins

料部

: Molten fuel pool↓

Unit-1：about 15 hours after scram

↑

燃

燃

部

Fuel zone

Fuel zone

↑

↓

↓

料

料
部

Unit-2: about 1 week after scram

Unit-2: about 109 hours after scram
Incorrect readings of water levelindicators considered in analysis

燃燃

Fuel zone↑

Fuel zone

↑

↓

↓

料

料

部

部

Unit-3: about 96 hours after scram
Incorrect readings of water levelindicators considered in analysis

Unit-3: about 1 week after scram

Figure 2.1-1

Analysis results by MAAP (core conditions)15Note: At Unit-1, about 80 tons of water was
25

injected from early morning to daylight on
original inventory for comparison.

20

15

10

Unit-3

３
号
機

5

２
号
機

Unit-2

１
号
機
Unit-1

Decay heat (MW)
崩壊熱（MW）

March 12, 2011. This was added to the0

Figure 2.2-1

1000

800

600

400

Decay heat changes and timings of no water injection at each unit

図２． 崩壊熱量と除熱量の比較
図２． 崩壊熱量と除熱量の比較
1000
of decay heat (GJ)
Amount崩壊熱量(GJ)

崩壊熱量(GJ)

図１．崩壊熱（ORIGENコードによる）

30

ジルカロイ（顕熱+融解熱）
Zircaloy(sensible +latent heat)
ジルカロイ（顕熱+融解熱）
UO2（顕熱+融解熱）
UO2（顕熱+融解熱）
UO2 (sensible+latent heat)
構造材（顕熱+融解熱）
構造材（顕熱+融解熱）
Structures (sensible +latent heat)
注水 注水
崩壊熱
Injectedwater
崩壊熱
炉水（燃料下端以下）
Decay heat
炉水（燃料下端以下）
炉水（燃料下端以上）
Reactor water (below BAF)
炉水（燃料下端以上）

800

600

Reactor water (above BAF)

400

200

200
0
1

１号機
Unit-1
2

3

4

２号機
Unit-2
5

２号機
5

6

6

7

３号機
Unit-3
8

9

0
1

１号機
2
3

Figure 2.2-2

4

7

３号機
8

Comparison of decay heat and heat removal capacities

16

9

A: RPV flange
A:圧力容器フランジ
B: ベローシール
Seal bellows
B:

Being
stabilized
安定傾向
bottom

C: Feedwater
C：給水ノズル
nozzle
D: RPV bottom
D:圧力容器下部
Pedestal

ペデスタル

- HVH
E:E:D/W
D/W-HVH
bottomFigure 3.1-1

Temperature changes of Unit-1

A:圧力容器フランジ
A: RPV flange

Beingstabilized
安定傾向
B:
B:ベローシール
Seal bellows

bottomWater injected
via CSCS注水

C: Feedwater
C：給水ノズル

bottomnozzle
D: RPV bottom
D:圧力容器下部
Pedestal

ペデスタル

- HVH
E:E:D/W
D/W-HVH
bottomFigure 3.1-2

Temperature changes of Unit-2

A:圧力容器フランジ
A: RPV flange

Beingstabilized
安定傾向

B:
Seal bellows
C: ベローシール

bottomWater injected via CS

CS注水

C: Feedwater
B：給水ノズル

bottomnozzle
D:圧力容器下部
D: RPV bottom
Pedestal

ペデスタル

E:E:D/W
D/W-HVH- HVH

bottomFigure 3.1-3

Temperature changes of Unit-3

17

Tamb
Qncuh
RPV上部壁
RPV
upper head

Quh

QraduhUpper internal
structures
上部構造材

Qfuh

Qfs

Mfed
Tin
給水系から
Water
injected viaの注水system
feedwater

Qs
Water
injected via CS
CS系からの注水

Mcs
Tin

Qradu

Mgc

CS水による
Condensation
by CS凝縮
water-Δ Qcsα

Tamb

Qfsh

Qsw Qradsw
Qncsw
RPV側壁
RPV
sidewall

炉
心
シ
ュ
ラ
ウ
ド

Qsh

Uncovered

露出炉心 Qc

Qradshcore

Qfc

Mgc,csα
Mglp

Qlp
冠水した燃料
Flooded fuelMfed+(Mcs-Mgc,csα )
-Mglp

熱の流れ
Heat
paths
水の流れ
Water
paths
図中の変数名は本文参照
See Attachment-4
or variables in the figure.

Figure 3.2-1

Heat balance model

18

Reference condensing
waterchamber
基準面器

Ｈｓ

Ｈｒ
Differential
pressure
gauge

差圧計

基準面器
側配管

Reference leg

Figure 3.3-1

炉側配管
Variableleg

Schematic illustration of a reactor water level indicator

Reference condensing
waterchamber
基準面器

Reference condensing
waterchamber
基準面器

基準面器
Apparent
water
から水位
level
increasesが減少すwhen
the waterlevel
in theることで、
reference
見かけ上condensing
の水位が
water
chamber上昇
decreased

Actualsituation
炉内水位
of reactor
water
の実際の
level
状況

Water level indicator readings
水位計の指示値
Differential

Differential

gaugepressure
差圧計
gauge

pressure
差圧計

Actual reactor water level
実際の水位

基準面器

Reference
leg
側配管

炉側配管

Variable leg

Figure 3.3-2

Reactor water level indicator readings when the water level in the
instrumentation line dropped

19

Pedestal
ペデスタル

About
約2.６m

About
約7.6m

Figure 4.1-1

ピット
Pits
(1.2m deep)(深さ1.2m)

Estimated fuel debris configuration when relocated to the PCV pedestal(fuel debris remained fluid and spread widely)Pedestal
ペデスタル

About
約2.６m

About
約7.6m

Figure 4.1-2

ピット
Pits
(1.2m deep)(深さ1.2m)

Estimated fuel debris configuration when relocated to the PCV pedestal(fuel debris clogged the pits)

20

Unit-1 R/B survey results

May 14, 2011
Tokyo Electric Power Company
Survey date: 16:01 to 17:39 on May 13, 2011 (Dose rates shown in green)
11:32 to 11:58 on May 5, 2011 (Dose rates shown in blue)
72(Two steps down from the top of the stairs)
Unit
04:18 to 04:47 on May 9, 2011 (Dose rates shown in red)79(Top of the stairs)
Floors 1 to 2 platform(Top of the stairs)

Above the platform 5.0
Below the platform 11.5

(Measured with inner door closed)

RCW piping
RCW piping

R/B large equipment
entrance hatch

Figure 4.2-1

Unit-1 R/B dose rate survey results

RCW
heat
ＲＣＷ
exchanger
熱交換器

床ドレン
サンプピット

機器ドレン
Equipment drain sump pit
サンプピットFloor drain sump pit

Figure 4.2-2

Schematic of interface between RCW and equipment drain pit

21Time after the accident (d)Figure 4.3-1

Evaluation of concrete erosion depth due to fuel debris relocated to the PCV

22

1u
120

30

110

VESSEL frange

Water injection
ratechanged
注水量変更
10/28 16:10
10/29 15:30
10/30 15:05
10/31 14:59
11/18 15:33

90

温度(℃)

80

RPVベローシール
RPV Seal bellows(HVH-12A)
Feedwater nozzle N4B (tip end)
給水ノズルN4B(終端)

3.8→4.5
4.5→5.5
5.5→6.5
6.5→7.5
7.5→5.5

20

70
60

15

50
40

10

30

Vessel Steam25

注水量(m^3/h)100

vessel core

Safetyvalve203-4A
ventilation 203-4A
安全弁排気
CRDハウジング上部
Housing upper part
CRDハウジング上部
Housing upper partvessel bottom head
(OP:15500)
Return air
D/W HVH戻り空気(HVH-12C)
S/C温度A①
Temperature A①

Temperature B①
S/C温度B①

Amount of water injected
注水量

20

5

10
0

0

6/27 循環注水冷却開Cooling by injected water
circulation
始・停止 started up/stopped
循環注水冷却停止
N2停止(6/27 8:0015:00)
地震

11/ 22

11/ 15

11/ 8

11/ 1

10/ 25

10/ 18

注水量3.5→4（高台炉
注水ポンプに切替）
注水量4→3.8
注水量3.3→3.8
注水量3.4→3.6
注水量3.5→3.7

Figure 5.1-1

注水量
Latest temperature changes of Unit-1 (October 18 to November 22,
2011)
大物搬入口前流量23

Figure 5.1-2

Steam blowout from floor penetration on Floor 1 filmed on June 3, 2011

Figure 5.1-3

No steam blowout from floor penetration on Floor 1 filmed on October 13,
2011

24

30
25
20
15
10

S/C温度
S/C Temperature

5
Amountof water injected
注水量

9/ 1

9/ 8

Figure 5.1-4

9/ 15 9/ 22 9/ 29

0
10/ 6 10/ 13 10/ 20 10/ 27 11/ 3 11/ 10

Temperature changes after increased water injection25

注水量（m3/h）

温度(℃)

VESSEL frange
Vessel Steam
RPV
Seal bellows
RPVベローシール
Feedwater
nozzle N4B (tip end)
給水ノズルN4B(終端)vessel core
Safety valve ventilation
安全弁排気203-4A 203-4A
Housing upper part
CRDハウジング上部
Housing upper part
CRDハウジング上部
vessel bottom head
Return air (HVH-12C)
D/W HVH戻り空気(HVH-12C)
Temperature A①
S/C温度A①
Temperature B①
S/C温度B①

150
140
130
120
110
100
90
80
70
60
50
40
30
20
10
0

140

D/W圧力
D/W pressure
Nitrogen gas filling
窒素注入圧力pressure135

圧力(kPa[abs])

130

125

120

115

110
10/19 10/21 10/23 10/25 10/27 10/29 10/31 11/2

Figure 5.1-5

11/4 11/6

11/8 11/10 11/12 11/14 11/16 11/18 11/20 11/22

Changes of D/W pressure and nitrogen gas filling pressure

26

Changes of N2 gas
injection amountcorresponding to
水素濃度上昇に伴うN2封入量
hydrogenconcentration
変更
increase
10/29 18:10
14→16.5Ｎｍ3/h

190
180

10/30 18:10 16.5→21Ｎｍ3/h
11/3 16:50 21→26.5Ｎｍ3/h

170
160
150
140
130

温度(℃)

120
110

Changes of water injection
corresponding totemperature increase20

温度上昇に伴う注水量変更
9/14 14:59
FDW3.8m3/h,CS1.0m3/h
9/15 15:45
FDW3.8m3/h,CS2.0m3/h
9/16 15:35
FDW3.8m3/h,CS3.0m3/h
9/19 15:35
FDW3.8m3/h,CS4.0m3/h
9/22 15:36
FDW3.8m3/h,CS5.0m3/h
9/26 15:05
FDW3.8m3/h,CS6.0m3/h
10/4 15:05
FDW3.8m3/h,CS7.0m3/h
11/18 15:05
FDW3.1m3/h,CS7.1m3/h

Feedwater nozzle N-4B
temperatureN-4B 温度
給水ノズル
69D2 OP.26601
Relief valve thermometerreadings
逃し安全弁漏洩検出器
RV-2-71A OP.17400

18

Main steam isolation valve
thermometerreadings
主蒸気隔離弁漏洩検出器
2-86A OP.12900

16

RPV bottom head upper side
temperature
RPV底部ヘッド 上部温度
69H1 OP.17232
RPV support skirt upper part
RPV支持スカート上部温度
temperature
69K1 OP.15394

14
12

100

10

90
80

8

70

注水量（m^3/h)

2
u 200

CRD housing upper parttemperature
CRDハウジング 上部温度
69N1 OP. 14175
RPV drain pipe upper part
RPVドレンパイプ
上部温度
temperature
106
106OP.11400近傍
Around OP.11400
D/W HVH Return air
D/W
HVH戻り空気(HVH(HVH-16A)
OP.10750
16A) OP. 10750
Reactor control room gas
temperature
原子炉抑制室ガス温度
114S
S/C Temperature A①
S/C温度A①

60

6

S/C
Temperature B①
S/C温度B①

50
40

4

30

Amount of注水量
water injected

20

2

10
0

0

11/22

11/15

11/8

11/1

10/25

10/18

Figure 5.2-1

Latest temperature changes of Unit-2 (October 18 to November 22, 2011)27

Figure 5.2-2

Steam blowout from immediately above the reactor on Floor 5 filmed on
September 17, 201128

Figure 5.2-3

No steam blowout from immediately above the reactor on Floor 5
filmed on October 20, 2011

29

3u
140

Changes of water injection
corresponding to
温度上昇に伴う注水量変更
temperature increase
27 11/18 15:33 FDW2.5 CS8.1

25

RPV
67A2 67A2
RPVｽﾀｯﾄﾞﾎﾞﾙﾄ温度
stud bolt temperature

24

RPV胴ﾌﾗﾝｼﾞ
body flange
RPV
69A269A2
RPV Seal bellows114L
114L
原子炉ﾍﾞﾛｰｼｰﾙ部

23

120

20

RPV胴ﾌﾗﾝｼﾞ下部温度
body flange lower
part
RPV
69B2
temperature 69B2
RPV
N4Bnozzle
69D1 N4B 69D1
RPV給水ﾉｽﾞﾙ
Feedwater
RPVbottom head upper
RPV 底部ﾍｯﾄﾞ上部
69H1
temperature 69H1
RPV
RPV下部ﾍｯﾄﾞ温度
bottom head 69L1
temperature 69L1

19

D/WHVH戻り空気
HVH 69L1 A号機
Return114A
air 114A
D/W

18

Water injection 7→9
注水量７→９17給水ラインからの炉注水3m3/hを追加給水ライン6m3/h、FPライン6m3/hに変更
給水ライン6m3/h、FPライン9m3/hに変更
給水ライン9m3/h、FPライン9m3/hに変更
給水ライン12m3/h、FPライン9m3/hに変更
給水ライン12m3/h、FPライン6m3/hに変更
給水ライン12m3/h、FPライン3m3/hに変更
給水ライン13.5m3/h、FPライン
2m3/hに変更
給水ライン13.5m3/h、FPライン
1m3/hに変更
給水ライン13.5m3/h、FPライン
0m3/hに変更
給水ライン12.5m3/h、FPライン
0m3/hに変更
給水ライン11.5m3/h、FPライン
0m3/hに変更
給水ライン10m3/h

22

80
温度(℃)

16
15
14

60

13
12
10.4

40

11

10.5
注水量
Amountof water injected

10.9

10
9

S/C S/C温度
temperature

8

20

7
7/14 20:01
N2 gas injection started up 14m3/h
7/14 20:01 N2封入開始 14m3/h

0

注水量（m^3/h)

21
100

6

給水ライン9.5m3/h

5給水ライン9m3/h

11/22

11/15

11/8

11/1

10/25

10/18

6/27 循環注水冷却開始・停止
循環注水冷却再開
循環注水冷却停止
S/C温度

Figure 5.3-1

Latest temperature changes of Unit-3 (October 18 to November 22, 2011)
N2封入開始給水ライン8m3/h
給水ライン7m3/h
給水ライン7m3/h, CS 1m3/h
給水ライン7m3/h, CS 2m3/h
給水ライン7m3/h, CS 3m3/h
給水ライン6m3/h, CS 3m3/h
給水ライン5m3/h, CS 3m3/h
給水ライン4m3/h, CS 3m3/h
給水ライン4m3/h, CS 8m3/h
給水ライン3m3/h, CS 8m3/h
給水ライン2.5m3/h, CS 8.1m3/h
注水量30

Spent fuel pool 62℃

Steam
blowout
experienced
beforeImmediately above RPV： 128℃
Figure 5.3-2

Temperature distribution of Unit-3 R/B filmed on March 20, 2011(by the SDF)

Steam blowout experienced before

Temporary equipment
storage pit

ReactorSpent fuel pool

Figure 5.3-3

Temperature distribution of Unit-3 R/B filmed on October 14, 201131

D/W pressure・S/C pressure

D/W pressure
S/C pressure

Date and time

Figure 5.3-4

Changes of D/W pressure and S/C pressure

32

Nitrogeninjection
Feedwater
systemlevel

Figure 6.1-1

Estimated conditions of the Unit-1 core

33

Core
spray
systemFigure 6.2-1

Feedwater
systemEstimated conditions of the Unit-2 core

34

Core
spray
syste
m

Feedwater
systemFigure 6.3-1

Estimated conditions of the Unit-3 core35

AttachmentsiAttachments

Attachment-1

Estimated core conditions disclosed in May 2011

Attachment-2

Core conditions estimated from heat balance while water injection was
being interrupted

Attachment-3

Core conditions estimated from the measured temperatures andpressures

Attachment-4

Core fuel temperatures estimated by the in-RPV temperature evaluation
model

Attachment-5

Calibration of water level indicators

Attachment-6

Radioactivity concentrations in the atmosphere of the containment vessel

Attachment-7

Operability checks of local power range monitor (LPRM) detectors (Unit-2,
Unit-3)(Attempt to estimate core conditions from LPRM data)

Attachment-8

Operability checks of control rod position indicator probes (PIPs) (Unit-1,
Unit-3)(Attempt to estimate core conditions from control rod position indicator
probe (PIP) data)

Attachment-9

Results of condition checks and behavior of drywell (D/W) sumpthermometers

Attachment-10

Behavior of the primary loop recirculation (PLR) pump inlet thermometers

Attachment-11

Contamination of the reactor building closed cooling water (RCW)
systemAttachment-12

Impacts of core-concrete reactions on the reactor containment vessel

Attachment-13

Estimation of the conditions of structural materials in the Unit-1
containment vessel

Attachment-14

Results of in-containment gas composition analysis

Attachment-15

Gas residues from the early phase of core-concrete reactions

Attachment-16

Paint stripping-off incidents of Unit-2 reactor building ceiling cranesiiAttachment-1
Estimated core conditions disclosed in May 2011

1.

Introduction
The Tohoku District Off-Pacific Ocean Earthquake (Great East Japan Earthquake) which

occurred on March 11, 2011 with its hypocenter off the Sanriku Coast, led to a series of
events in Unit-1 to Unit-3 of the Fukushima Daiichi Nuclear Power Station that caused a
severe accident as the units remained in continued station blackout conditions. The accident
far exceeded design basis accidents and even multiple accidents which had been assumed
in developing the accident management procedures, i.e., all emergency core coolingsystems could not function or were stopped including those of adjacent units.
In response to the request from the Nuclear and Industrial Safety Agency (NISA), theMinistry of Economy, Trade and Industry (METI) (*) received on April 25, 2011, plant data at
the time of the earthquake were collected to the extent possible and were reported to METI
on May 16, 2011. The Modular Accident Analysis Program (hereafter referred to as “MAAP”)
was applied to the system conditions, operation procedures and other parameters at thetime of the early stage of earthquake and evaluated plant conditions. The data acquired
were analyzed and the results were disclosed on May 23, 2011.
It should be noted, however, that the results disclosed had been based on the information
available at that time as well as the estimations and assumptions for the analysis. The
uncertainties of the results were quite large. The plant conditions could have turned out
significantly different from the analysis results with the progress of investigations thereafter.
Estimated core conditions disclosed on May 23, 2011, were based on a comprehensive
judgement derived from core conditions predicted by the MAAP analysis as well as fromtemperature behavior observed. Core conditions of Unit-1 to Unit-3 estimated and disclosed
on May 23, 2011, are summarized below.
(*)

Request of reports pursuant to Clause 1, Article 67 of the Reactor Regulation Law(Law No. 67) (Notification Number: H23-04-24 Gen-1), April 25, 2011, NISA, METI

2.

Unit-1, Fukushima Daiichi Nuclear Power Station

2.1.

MAAP analysis conditions

Table 1 gives key plant conditions for the analysis and Table 2 describes key event
sequences.
The following assumptions were made in the analysis, concerning the isolation
condensers (IC) and leaks from the containment vessel (PCV).

Attachment 1-1

①

Leaks from the PCV
Leaks from the PCV, specifically the drywell (D/W)), gaseous phase (about φ 3cm)

were assumed at about 18 hours after the earthquake. This assumption was made tosimulate the PCV pressure behavior actually measured. Enlarged leaks (about φ 7cm)
were also assumed at about 50 hours after the earthquake.
It should be noted that these assumptions were purely for analysis. It was not known
then (in May 2011) whether leaks from the PCV (D/W) had actually occurred, orwhether there had been inconsistencies between measured values and analysis due toproblems of measurement devices.
②

Operability conditions of isolation condensers
Operability of isolation condensers (ICs) was not assumed in the analysis, since theiroperability was not certain yet, after the unit had entered the station blackout (*). But
another situation was also analyzed, as a sensitivity analysis, in which the IC operability
was assumed for a limited time after the station blackout had started.
It should be mentioned that, in this sensitivity analysis, one IC sub-system wasassumed to have been operated intermittently during the time after the station blackout,on the grounds that the reactor pressures before the station blackout had beenchanging below the safety relief valve (SRV) actuation pressure (about 7.4MPa[abs]).

(*)

When the local water level indicators in the IC shell side were checked on October18, 2011, the readings were 65% in Sub-system A and 85% in Sub-system B (normal
level 80%).
According to the records on the IC coolant temperature chart, the temperature of the
Sub-system B stopped increasing at about 70 deg C. Coolant evaporation, whichcauses coolant water level changes, is considered to have been limited. The
temperatures of Sub-system A increased to about 100 deg C, saturation temperature,at around the time when tsunami arrived at the unit. The coolant water level decreaseof Sub-system A is considered to have been mainly due to heat exchange after thetsunami arrival.
It should be noted, however, that it is unknown, from the following reasons, to what
extent Sub-system A was operable and how long it was actually functioning after thetsunami arrival: ① the isolation valve aperture inside the PCV is unknown; ② the IC
heat removal performance deteriorates by accumulation of non-condensablehydrogen gas generated in water-zirconium reactions due to overheated fuel; and ③
the IC heat removal performance also deteriorates by the decreased amount of steam
Attachment 1-2

inflow to the IC from the reactor, because the reactor pressure dropped at the latest by
03:00 on March 12, 2011, and this caused the decrease of steam inflow to the IC.
In conclusion, the assumption in May 2011 of no IC operability after the station
blackout is considered to have been appropriate.

Table 1 Unit-1 key plant conditions
Item

Condition

Initial reactor power

1380MWt (rated)

Initial reactor pressure

7.03MPa[abs] (normal operating pressure)

Initial reactor water level

Normal level

Nodalization in RPV core

See Figure 4 in the reference (Outline of MAAP at
the end of this document)

Effective core nodalization

Radial: 5 nodes
Axial: 10 nodes

Cladding damage temperature

1000K

Core node melting point

2500K

PCV model

See Figure 5 in the reference (Outline of MAAP)

PCV free volume

D/W: 3410m3
S/C: 2620m3

Water volume in S/C pool

1750m3

Decay heat

Model ANSI/ANS5.1-1979(Parameters adjusted to simulate the ORIGEN2
decay heat with fuel burnup considered)

Attachment 1-3

Table 2

Unit-1 event sequence

Note
Conditions for analysis
No

Day & Time

1

3/11

2

3

Events to analyze

14:46

Earthquake

14:46

Reactor scrammed

14:47

MSIV closed

4

14:52

IC(A) (B) automatic start-up

5

About

IC(A) stopped

15:03
6

About

IC(B) stopped

15:03
7

8

9

10

15:17

15:19

15:24

15:26

IC(A) restarted

IC(A) stopped

IC(A) restarted

IC(A) stopped○: Recorded

Type

Remarks

○

－

○
○
○
△
△
△
△
△
△

△: Estimated from records□: Assumed in analysis○: Where records to refer to can be found△ or □: Grounds for estimation or assumption

4. Operation sheets, Supervisor shift transfer records in the 5/16
Report (*4)
4. Operation sheets, Supervisor shift transfer records in the 5/16
Report3. Alarm records by alarm loggers in the 5/16 Report
6. IC assumed closed from data recorded in the transient recorder inthe 5/16 Report
6. IC assumed closed from data recorded in the transient recorder inthe 5/16 Report
IC operation estimated from reactor pressure changes (2. Chart
records in the 5/16 Report) *1
IC operation estimated from reactor pressure changes (2. Chart
records in the 5/16 Report) *1
IC operation estimated from reactor pressure changes (2. Chart
records in the 5/16 Report) *1
IC operation estimated from reactor pressure changes (2. Chart
records in the 5/16 Report) *1

11

12

13

14

15:32

15:34

15:37

18:10

IC(A) restarted

IC(A) stopped

Station blackout

IC(A) Valve-2A and 3A opened/Steamgeneration confirmed

15

16

18:25

21:19

IC(A) Valve-3A closed

IC connected to diesel-driven fire pump(D/D-FP)

17

18

20

21:30

21:35

05:46

IC Valve-3A opened

IC in service by D/D-FP input

Freshwater injection started by fireengines

21

14:30

△
△
○
□
□
□
□
□

○

PCV pressure decrease confirmed byactuating AO valve on suppression
chamber (S/C) side at 10:17 (PCV
venting)

IC operation estimated from reactor pressure changes (2. Chart
records in the 5/16 Report) *1
IC operation estimated from reactor pressure changes (2. Chart
records in the 5/16 Report) *1
4. Operation sheets, Supervisor shift transfer records in the 5/16
ReportIC functions assumed lost after station blackout, although this part ismentioned in 7. Operational procedure records in the 5/16 Report *2
IC functions assumed lost after station blackout, although this part ismentioned in 7. Operational procedure records in the 5/16 Report *2
IC functions assumed lost after station blackout, although this part ismentioned in 7. Operational procedure records in the 5/16 Report *2
IC functions assumed lost after station blackout, although this part ismentioned in 7. Operational procedure records in the 5/16 Report *2
IC functions assumed lost after station blackout, although this part ismentioned in 7. Operational procedure records in the 5/16 Report *2

7. Operational procedure records in the 5/16 Report *3

7. Operational procedure records in the 5/16 Report, PCV venting△

assumed done at 14:30 when the PCV pressure decrease wasconfirmed

22

14:49

PCV vent valve closed△

PCV vent valve closure assumed from PCV pressure increase

23

14:53

Freshwater injection ended○

7. Operational procedure records in the 5/16 Report

24

15:36

Explosion at Unit-1 reactor building○

7. Operational procedure records in the 5/16 Report

25

20:20

Seawater injection started○

7. Operational procedure records in the 5/16 Report *3*1 The IC operability prior to the station blackout was not well known, but in the analysis one of two sub-systems was assumed to have been in
operation intermittently, because the reactor pressures were changing between about 6.2 to 7.2MPa[abs], according to the records on thecharts (Item 2 in the May 16 Report), while the preset actuation pressure for relief functions of SRV No. 1 was about 7.4MPa[abs] and that for
stopping the blow-out was about 6.9MPa[abs].
*2 The IC operability after the station blackout started was not well known, either. In the analysis, the ICs were assumed to have lost theirfunctions, because the records of IC operation were insufficient.
*3 Timings of changing the water injection flow rates and their amounts were set so as not to exceed the daily average flow rates and the total
injection amount, based on the day-to-day records of injection amount to the reactor (Item 7. Compilation of operational procedures in the
May 16 Report).
*4 Operation records, analyses of accident records and evaluation of their impacts at the Fukushima Daiichi Nuclear Power Station at the timeof Tohoku District Off-Pacific Ocean Earthquake, May 16, 2011, Tokyo Electric Power Company

2.2.

Results of MAAP analysis
Table 3 summarizes the analysis results.

Table 3 Summary of Unit-1 analysis results
Item

Results

Timing of core starting to be uncovered

About 3 hours after the earthquake

Timing of core starting to be damaged

About 4 hours after the earthquake

Timing of RPV being damaged

About 15 hours after the earthquake

Details of the results follow.
The reactor water level decreased to the level of top of active fuel (TAF) in about
2 hours after the IC was assumed to have stopped its functions after the tsunami
arrival and thereafter, the core was damaged (Figure 1).
The reactor water level actually measured changed in the core region after theearthquake. This was quite different from the analysis results. In the analysis, thewater level could not be maintained in the reactor pressure vessel (RPV) and the
RPV was damaged. It is possible that the water in the water level indicators
evaporated due to elevated PCV temperatures and the water level indicatorreadings were not correct. As a matter of fact, water level indicators of Unit-1 were
calibrated at a later date by water filling and the reactor water level was found tohave been below the core.
The RPV pressure increase after IC shutdown was assumed but this wasmaintained at about 8MPa by the SRV functions. After the core damage, fueldebris was relocated to the lower plenum, and about 15 hours after the earthquake
the RPV was damaged and the reactor pressure quickly dropped (Figure 2).
The PCV pressures increased temporarily due to the steam and hydrogen gas(generated by the water-zirconium reactions in the core) discharged from the RPV
but then decreased due to the leaks assumed from the PCV and later on March 12,
2011, they sharply dropped due to venting (Figure 3).
Water injection to the reactor started about 14 hours after the IC shutdown was
assumed. By that time, the fuel was molten due to the decay heat, had relocated to
the lower plenum, and the RPV was damaged about 15 hours after the earthquake(Figure 4).

Attachment 1-7

2.3.

Estimation of Unit-1 core conditions

The core conditions were estimated as follows, by comprehensively considering theinformation available including the core conditions being estimated from the temperature
behavior actually measured.
The analysis gave the results that the core damage had started relatively early after theunit had entered station blackout and that the RPV damage had followed. The results wereconsidered to have predicted the accident progression as being severer than the actual
progression, when the plant conditions, estimated from the following temperatures atvarious points of the reactor and other information, were considered.
When temperatures became measurable at various points of the reactor, the RPV
temperatures were above 400 deg C at multiple measurement points. At that timing, thecore was considered to have been under insufficiently cooled situations. But, as soon as
water injection to the reactor was certainly ensured by using the feedwater (FW) line, theRPV temperatures decreased quickly. The core was considered then to have been
undergoing cooling sufficiently.
Most fuel was considered as being cooled in the RPV from the following observations(Figure 5): temperatures under the RPV, e.g., those of CRD housings, had been measurable
but they would not have been measurable if the RPV had been broken; the RPV steel
temperatures were changing in the region of 100 to 120 deg C, and at multiple
measurement points the temperatures responded in a consistent manner to the changing
amount of water injection; and the temperatures in the upper part of the RPV were in higher
trends at multiple points, indicating the heat source was present in the RPV. When theanalysis was made, MAAP was considered to have predicted the severer accident
progression than the reality, on the grounds that the ICs were known to have been out ofservice at the time of tsunami arrival, but the detailed operation conditions were unclear.
Consequently, based on the analysis results and plant parameters (temperatures aroundthe RPV), the core was considered to have been significantly damaged and relocated to alocation lower than the original position (lower plenum), and it was being cooled there in astable manner.

Attachment 1-8

10
Shroud water level (analysis)
シュラウド内水位（解析）8

Downcomer waterダウンカマ水位（解析）level (analysis)

TAF到達(約3時間後)
Loweredto TAF (about 3 hours after earthquake)
6

Actual water level実機計測値（燃料域A）
indicator reading (Fuel range A)
Actual water level実機計測値（燃料域B）
indicator reading (Fuel range B)

BAF到達(約5時間後)

Lowered to BAF (about 5 hours after earthquake))※：Water level (analysis) after the RPV

4damage does not mean the waterlevel was being maintained.

原子炉水位（m）

Water
injection started (about 15 hours after earthquake)
注水開始(約15時間後)
2

TAF

0

-2
BAF
-4

-6

-8

-10
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

3/16
0:00

3/16
12:00

Date日時and time

Figure 1

Unit-1 reactor water level changes

10
RPV pressure (analysis)

RPV破損（約15時間後）
RPV
damaged (about 15 hours after earthquake)

RPV圧力（解析）readings
Actual pressure
indicator(Sub-system A)
実機計測値（ A系）

Actual pressure indicator readings
実機計測値（B系）(Sub-system
B

原子炉圧力(MPa[abs])

8

6

Pressure
decrease
IC起動による
圧力低下

by IC start-up
4

2

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

日時timeDate andFigure 2

Unit-1 RPV pressure changes

Attachment 1-9

3/15
12:00

3/16
0:00

3/16
12:00

1.6

1.4

Leaks from PCV
assumed (about 18 hours after earthquake and later)
格納容器漏えいを仮定(約18時間後以降)

D/W pressures (analysis)
D/W圧力（ 解析）S/C pressures(analysis)
S/C圧力（解析）
実機計測値（D/W）
Actual pressure
indicator readings(D/W)
実機計測値（ S/C）
Actual pressure indicator readings (S/C)

原子炉格納容器圧力(Mpa[abs])

1.2
S/Cベント
S/C
vent

格納容器漏えいを仮定(約50時間後以降)
Leaks from
PCV assumed (about 50 hours after earthquake and later)

1.0

0.8

0.6

0.4

0.2
RPV破損(約15時間後)

RPV damaged (about 15 hours after earthquake)
0.0
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

日時timeDate andFigure 3

Unit-1 PCV pressure changes

Attachment 1-10

0:00

3/15
12:00

3/16

0:00

↑

↑

燃

燃

料

料

部

部

↓

↓

About 4.7 hours after scram

About 5.3 hours after scram

↑

↑

燃

燃

料

料

部

部

↓

↓

About 15 hours after scram

About 14.3 hours after scramCore damage schematic: No fuel (collapsed): Intact fuel: Sedimented damaged fuel (pin shape was kept): Fuel pin diameters enlarged by solidified molten fuelwhile going down pin surfaces: Coolant flow paths blocked by enlarged fuel pins
: Molten fuel pool
Figure 4

Unit-1 estimated core conditions

Attachment 1-11

425.0【留意事項】[Note]

400.0

各計測器については、地震やその後の事象進展の影響を受けて、通常の使用環境条件を超えているもVesselVESSELflange
frangeのもあり、正しく測定されていない可能性のある計測器も存在している。プラントの状況を把握する
Some
devices may give incorrect readings in the working

375.0

ために、このような計器の不確かさも考慮したうえで、複数の計測器から得られる情報を使用して変化の傾向にも着目して総合的に判断している。
environment
after the earthquake and in the accident progression

350.0

thereafter as they were operating beyond their normal duty

VesselVessel flange
frangeconditions. Readings from multiple devices are interpreted

325.0

comprehensively in estimating the plant conditions, with their

RPVseal bellows
RPVベローシール(HVH(HVH-12A)
12A)

300.0

uncertainties taken into consideration.

給水ノズルN4B（終端）Feedwater
nozzle N4P

275.0

vessel core

250.0

Safetyvalve exhaust
安全弁排気203-4A①

℃

225.0

S/Cへ

200.0

RPV
lower part
圧力容器下部D/W HVH
returnHVHline
D/W
HVH戻り（(HVH-12C)
12C）

175.0

CRD
housing upper part
CRDハウジング上部

150.0
125.0

CRDハウジング下部
CRD housing lower part

100.0

S/Cプール水温度AS/C
pool temperatures A

75.0

S/Cプール水温度B
S/C
pool temperatures B

50.0
25.0
0.0
3/
20

3/
22

3/
24

3/
26

3/
28

3/
30

4/ 1 4/ 3 4/ 5 4/ 7 4/ 9

Figure 5

4/
11

4/
13

4/
15

4/
17

4/
19

4/
21

4/
23

4/
25

4/
27

4/
29

5/ 1 5/ 3 5/ 5 5/ 7 5/ 9

5/
11

5/
13

5/
15

5/
17

Unit-1 temperature changes at typical points (May 2011 when disclosed)

3.

Unit-2, Fukushima Daiichi Nuclear Power Station

3.1.

MAAP analysis conditions
Table 4 gives key plant conditions for the analysis and Table 5 describes key
event sequences.
The following two cases were analyzed, and the following assumptions weremade concerning the leaks from the PCV.

① Two cases analyzed
At Unit-2, seawater injection started at 19:54 on March 14, 2011, as seen in Table 5.
The following two cases were analyzed concerning the amount of seawater injection
thereafter.
Case 1: The reactor water level actually measured (at around the core center) was
simulated by setting the amount of water injected to the reactor in the analysis
at lower values than those measured at the fire engine discharge point.
Case 2: The amount of water injected to the reactor in the analysis was set at a lower
value than those measured at the fire engine discharge point so that thereactor water level in the analysis could maintain the level approximately
below the core level in the reactor. This case was chosen, because the reactor
water level indicators might have failed to give correct readings, as had been
experienced at Unit-1, and the reactor water level might not have been
maintained in the core region.
② Leaks from the PCV
Leaks from the PCV, specifically the D/W, gaseous phase (about φ 10cm) wereassumed at about 21 hours after the earthquake. This assumption was made to simulate
the PCV pressure behavior actually measured. In addition, leaks from the PCV,
specifically the S/C), gaseous phase (about φ 10cm) were assumed at the timing when
unusual sounds around the S/C had been noticed on March 15, 2011.
It should be noted that these assumptions were purely for analysis. It was not known by
then (May 23, 2011) whether leaks from the PCV (D/W) had actually occurred, or whether
there had been inconsistencies between measured values and analysis due to problemsof measurement devices.

Attachment 1-13

Table 4

Unit-2 key plant conditions

Item

Condition

Initial reactor power

2381MWt (rated)

Initial reactor pressure

7.03MPa[abs] (normal operating pressure)

Initial reactor water level

Normal level

Node division in RPV core

See Figure 6 in the reference (Outline of MAAP)

Effective core node divisions

Radial: 5 nodes
Axial: 10 nodes

Cladding damage temperature

1000K

Core node melting point

2500K

PCV model

See Figure 7 in the reference (Outline of MAAP)

PCV open volume

D/W: 4240m3
S/C: 3160m3

Water volume in S/C pool

2980m3

Decay heat

Model ANSI/ANS5.1-1979(Burnup at the end of equilibrium core assumed)

Attachment 1-14

Table 5

Unit-2 event sequence

Note
Conditions for analysis
No

Day & Time

1

3/11

Events to analyze○: Recorded

Type

Remarks△: Estimated from records□: Assumed in analysis○: Where records to refer to can be found△ or □: Grounds for estimation or assumption

14:46

Earthquake

○

－

2

14:47

Reactor scrammed○

4. Operation sheets, shift supervisor logbooks in the 5/16 Report

3

15:02

RCIC manually started up○

7. Operational procedure records in the 5/16 Report

4

15:28

RCIC tripped (L-8)○

7. Operational procedure records in the 5/16 Report

5

15:41

Station blackout○

4. Operation sheets, shift supervisor logbooks in the 5/16 Report

04:20

RCIC water source switched from

6

3/12

7. Operational procedure records in the 5/16 Reportcondensate storage tank to S/C

○

13:25

RCIC stopped

○

16:34

RPV depressurization operation started

to
05:00
7
8

3/14

(opening SRV1)
16:34

Seawater injection started via fire○○

7. Operational procedure records in the 5/16 Report
7. Operational procedure records in the 5/16 Report

7. Operational procedure records in the 5/16 Report *1

engine water lines
9

About

RPV pressure decrease confirmed

18:00
10

19:20

Fire engines stopped pumping (out of
fuel)

11

19:54

First fire engine started pumping○○○

7. Operational procedure records in the 5/16 Report

7. Operational procedure records in the 5/16 Report *1

7. Operational procedure records in the 5/16 Report *1, *2

12

19:57

Second fire engine started pumping

21:20

RPV depressurized by opening SRV2
and reactor water level recovered

13

About

SRV1 closure assumed

23:00
14

3/15

About

Unusual sounds near the S/C, followed

06:14

by pressure decrease in the chamber

○
○
□
○

7. Operational procedure records in the 5/16 Report *1
7. Operational procedure records in the 5/16 Report *1

SRV1 closure assumed, as the RPV increased at this timing at about
23:00.
Press release (TEPCO HP (http://www.tepco.co.jp/index-j.html)*1 The seawater injection was assumed to have started at 19:54 on March 14, 2011, when the reactor water level increase had been
confirmed. There is an earlier record at 19:20 on March 14, 2011, “fire engines stopped,” which means that some amount of seawater may
have been injected till then after 16:34 on March 14, 2011, but the effect of this temporary water injection was ignored in the analysis.
*2 Timings of changing the water injection flow rates and their amount were set so as not to exceed the daily average flow rates and the total
injection amount, based on the day-to-day records of injection amount to the reactor (Item 7. Compilation of operational procedures in the
May 16 Report).

3.2.

Results of MAAP analysis

3.2.1. Results of Case 1 analysis
Table 6 summarizes the analysis results.

Table 6 Summary of Unit-2 analysis results Case 1
Item

Results

Timing of core starting to be uncovered

About 75 hours after the earthquake

Timing of core starting to be damaged

About 77 hours after the earthquake

Timing of RPV being damaged

No RVP damage occurred in this analysis

Details of the Case 1 analysis results follow.
The reactor water level gradually decreased after the RCIC had stopped, thecore started to be uncovered, the core became totally uncovered by the SRV
actuation, and the core started being damaged (Figure 6). Almost at the same time,
water injection was started, but the amount of water injected had been adjusted tosimulate the reactor water level indicated by the indicator readings. The amount
was insufficient and the water level just deep enough to cover only about half of thecore could be maintained. The core became damaged.
The RPV pressures were kept high at around the SRV actuation pressure until
the RCIC was shut down. Upon SRV actuation after the RCIC shutdown, thereactor was rapidly depressurized and pressure gradually decreased to around theatmospheric pressure.
While the RCIC was functioning, the measured reactor pressures changed at alower level than the analyzed pressure. There is a possibility that a leak path fromthe PCV to S/C via the SRVs was formed, but it was not known then (May 2011)
whether there were actual leaks or whether there were simply instrumentation
system problems. After SRV actuations, the reactor pressures changed roughly ina consistent manner in measurement and analysis (Figure 7).
The PCV pressures increased following the S/C pool temperature increase, but
the D/W pressure increasing rate after the earthquake until just before theassumed leaks slowed down, consistent with the measurement, since leaks fromthe PCV (D/W) had been assumed. Upon SRV actuation on March 14, 2011, thePCV pressures showed a temporary increase. Measured PCV pressures began to

Attachment1-17

decrease thereafter. In the analysis, too, leaks from the S/C gaseous phase wereassumed to have occurred when unusual sounds were noticed in the vicinity of the
S/C on March 15, 2011 (Figure 8).
The Unit-2 core was concluded to have remained in the core region, althoughpart of the molten fuel remained as a pool, and the RPV was concluded not to havebeen damaged. This was considered to have been feasible because, water
injection via the RCIC at an early stage could have been implemented fairly
continuously, and the time delay from the RCIC shutdown to water injection
initiation could have been shorter than at Unit-1 (Figure 9).

3.2.2. Results of Case 2 analysis
Table 6 summarizes the analysis results.

Table 7 Summary of Unit-2 analysis results Case 2
Item

Results

Timing of core starting to be uncovered

About 75 hours after the earthquake

Timing of core starting to be damaged

About 77 hours after the earthquake

Timing of RPV being damaged

About 109 hours after the earthquake

Details of the Case 2 analysis results follow.
The reactor water level gradually decreased after the RCIC had stopped, thecore started to be uncovered, the core became totally uncovered by the SRV
actuation, and the core started being damaged. Almost at the same time, water
injection was started, but the amount of water injected had been adjusted as
insufficient to maintain the reactor water level above the bottom of active fuel(Figure 10).
The RPV pressures showed a temporary increase after depressurization by the
SRV actuation. This was due to the steam produced when part of the molten fuelwas relocated to the lower plenum. But the overall behavior (Figure 11) other than
this point was similar to the behavior in Case 1.
The PCV pressures showed a temporary increase, as in the case of reactor
pressures, due to the steam produced when part of molten fuel was relocated to
the lower plenum. But the overall behavior (Figure 12) other than this point was
similar to the behavior in Case 1.

Attachment1-18

Part of the molten fuel remained in the RPV, but the RPV was damaged. The
amount of water injected had been set at a lower value than that in Case 1 and thiscaused the core damage to be advanced (Figure 13).

3.3.

Estimation of Unit-2 core conditions

The core conditions are estimated as follows, by comprehensively considering theinformation available including the core conditions being estimated from the temperature
behavior actually measured.
The analysis of Case 1 gave the results that the fuel in the core had been molten but
remained in the core region, although part of the molten fuel had remained as a pool, andthat the RPV had not been damaged. The analysis of Case 2 gave the results that part ofmolten fuel had remained in the RPV, but the RPV had been damaged.
Most of the fuel was considered to have been undergoing cooling in the RPV from thefollowing plant parameters observed: the RPV bottom temperatures were changing ataround 100 to 120 deg C and at more than one measurement points the temperatures
responded in a consistent manner to the changing amount of water injected; and the heatsource was considered to be in the RPV, because the RPV upper part gave higher
temperatures (Figure 14).
Consequently, based on the analysis results and plant parameters, the core wasconsidered to have been significantly damaged and relocated to a location lower than its
original position (lower plenum), and it was mostly being cooled there in a stable manner.

Attachment1-19

10

RCICRCIC停止
stopped8

Shroud waterlevel (analysis)
シュラウド内水位(解析）Downcomer
water level (analysis)
ダウンカマ水位(解析)

SRV開
SRV
opened

実機計測値（燃料域A)
Actual waterlevel indicator readings(Fuel range water level indicator A)

6

Lowered to TAF (about 75 hours after earthquake)

原子炉炉水位 (m)4

TAF到達(約75時間後)

BAF到達(約76時間後)
Loweredto BAF (about 76 hours after earthquake)

2
TAF

0
-2

RCIC起動
RCIC
started up
BAF

-4
-6
海水注水開始Seawater
injection started

-8
-10
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

3/16
0:00

3/16
12:00

3/17
0:00

3/17
12:00

3/18
0:00

3/18
12:00

Date 日時and time

Figure 6

Unit-2 reactor water level changes Case 1

10
SRV開opened
SRV
RCIC RCIC停止
stopped

RCIC
started up
RCIC起動RPV圧力(解析）RPV pressure (analysis)
Actual pressure実機計測値
indicator readings

原子炉圧力 (MPa[abs])

8

6

4

2
計装バッテリ枯渇
Hunting oscillation
due to depletionに伴うハンチング
of batteries for measurement
0

-2
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
3/15
3/15
12:00
0:00time 12:00
Date and

3/16
0:00

3/16
12:00

3/17
0:00

Date and
日時timeFigure 7

Unit-2 RPV pressure changes Case 1

Attachment1-20

3/17
12:00

3/18
0:00

3/18
12:00

1
D/W圧力(解析）SRVSRV開
opened0.8

原子炉格納容器圧力 (MPa[abs])

D/W pressures (analysis)
S/C圧力(解析)S/C pressures (analysis)
実機計測値(D/W)
実機計測値(S/C)
Actual pressure indicatorreadings (D/W)
Actual pressure indicator readings (S/C)

Leaks D/Wに漏えいを仮定
from D/W assumed(約21時間後）(about 21 hours
after earthquake)
0.6

0.4

0.2

0
S/C付近で異音：S/Cに漏えいを仮定(約87時間後)
Unusual sounds near S/C: Leaks from S/Cassumed (about 87 hours after earthquake)
-0.2
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

3/16
0:00

3/16
12:00

3/17
0:00

Date 日時and time

Figure 8

Unit-2 PCV pressure changes Case 1

Attachment1-21

3/17
12:00

3/18
0:00

3/18
12:00

↑

↑

燃

燃

料

料

部

部

↓

↓

About 87 hours after scram

About 96 hours after scram

↑

↑

燃

燃

料

料

部

部

↓

↓

About 120 hours after scram

About 1 week after scramCore damage schematic: No fuel (collapsed): Intact fuel: Sedimented damaged fuel (pin shape was kept): Fuel pin diameters enlarged by solidified molten fuelwhile going down pin surfaces: Coolant flow paths blocked by enlarged fuel pins
: Molten fuel pool
Figure 9

Unit-2 core conditions Case 1

Attachment1-22

10
Shroud waterlevel (analysis)
シュラウド内水位(解析）RCIC RCIC停止
stopped8

Actual waterlevel indicator readings
実機計測値（燃料域A)(Fuel range water level indicator A)

6

Lowered to TAF(about 75 hours
after earthquake)
TAF到達(約75時間後)

原子炉水位 (m)

2
TAF

0

※：Water levels (analysis) after

RPV damage (about 109
hours after the earthquake
Loweredto BAF
BAF到達(約76時間後)(Table 7)) does not mean the(about 76 hours
water level is maintained atafter earthquake)that level4-2

Downcomer
water level (analysis)
ダウンカマ水位(解析)

SRV開
SRV
openedRCIC
started up
RCIC起動BAF-4-6
海水注水開始Seawater injection
started

-8
-10
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

3/16
0:00

3/16
12:00

3/17
0:00

3/17
12:00

3/18
0:00

3/18
12:00

Date日時and time

Figure 10

Unit-2 reactor water level changes Case 2

10

SRVSRV開
openedRCIC
started up
RCIC起動RPV pressure RPV圧力(解析）(analysis)

RCIC stopped
RCIC停止Actual pressure実機計測値
indicator readings8

原子炉圧力 (MPa[abs])

下部プレナムへのリロケーション(解析)Fuel relocation
to lower plenum (analysis)
6

RPVdamaged (about 109 hours after earthquake)
RPV破損(約109時間後)
42
計装バッテリ枯渇
Hunting oscillation
due to depletionに伴うハンチング
of batteries for measurement

0

-2
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

3/16
0:00

3/16
12:00

3/17
0:00

Date
and time
日時

Figure 11

Unit-2 RPV pressure changes Case 2

Attachment1-23

3/17
12:00

3/18
0:00

3/18
12:00

1
D/W圧力(解析）D/W pressures (analysis)
S/C圧力(解析)S/C pressures (analysis)
実機計測値(D/W)
SRV
openedSRV開原子炉格納容器圧力 (MPa[abs])

0.8

実機計測値(S/C)
Actual pressure indicatorreadings (D/W)

Actual pressure indicator readings (S/C)

Leaks from D/W assumed
D/Wに漏えいを仮定(about 21(約21時間後）
hours after earthquake)
0.6

0.4

0.2

0

Unusual sounds near S/C: Leaks from S/C
S/C付近で異音：S/Cに漏えいを仮定(約87時間後)

assumed (about 87 hours after earthquake)

-0.2
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

3/16
0:00

3/16
12:00

3/17
0:00

Date 日時and time

Figure 12

Unit-2 PCV pressure changes Case 2

Attachment1-24

3/17
12:00

3/18
0:00

3/18
12:00

↑

↑

燃

燃

料

料

部

部

↓

↓

About 87 hours after scram

About 96 hours after scram

↑

↑

燃

燃

料

料

部

部

↓

↓

About 100 hours after scram

About 109 hours after scramCore damage schematic: No fuel (collapsed): Intact fuel: Sedimented damaged fuel (pin shape was kept): Fuel pin diameters enlarged by solidified molten fuelwhile going down pin surfaces: Coolant flow paths blocked by enlarged fuel pins
: Molten fuel pool
Figure 13

Unit-2 core conditions Case 2

Attachment1-25

1F-2 温度に関するパラメータ（代表点）[Note]

425

Some devices may give incorrect readings in the working【留意事項】
environment
after the earthquake and in the accident progression
各計測器については、地震やその後の事象進展の影響を受けて、通常の使用環境条件を超えているも

400

Feedwater
給水ノズルnozzle
N-4BN-4B
温度のもあり、正しく測定されていない可能性のある計測器も存在している。プラントの状況を把握する
thereafteras they were operating beyond their normal duty
ために、このような計器の不確かさも考慮したうえで、複数の計測器から得られる情報を使用して変375化の傾向にも着目して総合的に判断している。
conditions.
Readings from multiple devices are interpreted

Relief valve thermometer
逃し安全弁漏洩検出器
readings RV-71A
RV-2-71A

comprehensively in estimating the plant conditions, with their350

uncertainties taken into consideration.

325

Main steam isolation valve
主蒸気隔離弁漏洩検出器
thermometer readings 2-86A
2-86A

300

RPV
support skirt upper part
圧力容器支持スカート上
temperature
部温度

275

RPV drain pipe upper part
圧力容器ドレンパイプ
上temperature
部温度

225

D/W HVH
Return air
D/W
HVH戻り温度(HVH(HVH-16A)
16A)

℃

250

S/Cへ

200
175

CRD
housing upper
part
CRDハウジング
上部温度

125* Incorrect
readings in the box※実線枠は計器不良

S/Cプール水温度AS/C
pool temperatures A

100

S/C pool temperatures B
S/Cプール水温度B

75
50
25
3/ 20 3/ 23 3/ 26 3/ 29

temperature

RPV
lower part
圧力容器下部温度

150

0Reactor control room gas
原子炉抑制室ガス温度

4/ 1

4/ 4

4/ 7

4/ 10 4/ 13 4/ 16 4/ 19 4/ 22 4/ 25 4/ 28

Figure 14

5/ 1

5/ 4

5/ 7

5/ 10 5/ 13 5/ 16 5/ 19 5/ 22 5/ 25 5/ 28

Unit-2 temperature changes at typical points (May 2011 when disclosed)

4.
4.1.

Unit-3, Fukushima Daiichi Nuclear Power Station
MAAP analysis conditions
Table 8 gives key plant conditions for the analysis and Table 9 describes key
event sequences.
The following two cases were analyzed.

At Unit-3, fresh water injection started at 09:25 on March 13, 2011, as Table 9 shows.
The following two cases of water injection amount thereafter were analyzed.
Case 1: The reactor water levels actually measured (at around the core center) were
simulated by setting the amount of water injected to the reactor in the analysis at
the lower values than those measured at the fire engine discharge point.
Case 2: The amount of water injected to the reactor in the analysis was set at lower
values than those measured at the fire engine discharge point so that the reactor
water level in the analysis could maintain the level approximately below the corelevel in the reactor. This case was chosen, because the reactor water levelindicators might have failed to give correct readings, as had been experienced at
Unit-1, and the reactor water level might not have been maintained in the coreregion.

Table 8Unit-3 key plant conditions

Item

Condition

Initial reactor power

2381MWt (rated)

Initial reactor pressure

7.03MPa[abs] (normal operating pressure)

Initial reactor water level

Normal level

Node division in RPV core

See Figure 6 in the reference (Outline of MAAP)

Effective core node divisions

Radial: 5 nodes
Axial: 10 nodes

Cladding damage temperature

1000K

Core node melting point

2500K

PCV model

See Figure 7 in the reference (Outline of MAAP)

PCV open volume

D/W: 4240m3
S/C: 3160m3

Water volume in S/C pool

2980m3

Decay heat

Model ANSI/ANS5.1-1979(Burnup at the end of equilibrium core assumed)

Attachement1-27

Table 9

Unit-3 event sequences

Note○: Recorded

Conditions for analysis
No

Day & Time

1

3/11

Type

Remarks

Events to analyze△: Estimated from records□: Assumed in analysis○: Where records to refer to can be found△ or □: Grounds for estimation or assumption

14:46

Earthquake

○

－

2

14:47

Reactor scrammed○

4. Operation sheets, shift supervisor logbooks in the 5/16 Report

3

15:06

RCIC manually started up○

7. Operational procedure records in the 5/16 Report

4

15:25

RCIC tripped (L-8)○

7. Operational procedure records in the 5/16 Report

5

15:38

Station blackout○

4. Operation sheets, shift supervisor logbooks in the 5/16 Report

6

16:03

RCIC manually started up○

7. Operational procedure records in the 5/16 Report

11:36

RCIC tripped○

7. Operational procedure records in the 5/16 Report

12:35

HPCI started up (L-2)○

7. Operational procedure records in the 5/16 Report

02:42

HPCI stopped○

7. Operational procedure records in the 5/16 Report

About

RPV depressurization operation by

09:08

SRVs

09:20

PCV

7

3/12

8
9
10

11

3/13

pressure

decrease

by

○

PCV

7. Operational procedure records in the 5/16 Report

PCV venting assumed to have started at this timing 09:20, when PCV

venting confirmed

pressure○

decrease

wasconfirmed,

although

the

vent

line

configuration is recorded to have been completed at 08:41 byactuating the S/C side AO valve in 7. Operational procedure records
in the 5/16 Report

○

12

09:25

Freshwater injection started

13

11:17

AO valve closure on the PCV vent line
confirmed due to pressure loss for

7. Operational procedure records in the 5/16 Report *1
7. Operational procedure records in the 5/16 Report○driving air cylinder

14

12:30

Valve actuated on the PCV vent line

15

13:12

Freshwater

injection

switched○toseawater injection
16

14:10

○

Valve closure assumed on the PCVvent line

7. Operational procedure records in the 5/16 Report
7. Operational procedure records in the 5/16 Report *1

PCV venting, which started at 12:30 on 3/13 (No.14), assumed to have△ended at this timing 14:10 from D/W pressure increase. Note: 7.
Records of operational procedures in the 5/16 Report records thevalve closure was confirmed at 16:00 on 3/15.

17

3/14

01:10

Water injection halted to allow filling for
the water source pit

18

03:20

Water source pit filled, reactor water
injection restarted

19

05:20

S/C side AO valve actuated for PCV○○○

7. Operational procedure records in the 5/16 Report

7. Operational procedure records in the 5/16 Report *1

7. Operational procedure records in the 5/16 Report

venting
20

12:00

S/C side valve closure assumed for
PCV venting

PCV venting, which started at 05:20 on 3/14 (No.19), assumed to have△ended at this timing 12:00 from D/W pressure increase. Note: 7.
Operational procedure records in the 5/16 Report records the valve
closure was confirmed at 16:00 on 3/15

21

16:00

S/C side valve actuation assumed for△

PCV venting assumed at this timing from D/W pressure decrease

PCV venting
22

21:04

S/C

side

valve

closureoperation

assumed for PCV venting
23

3/15

16:05

S/C side valve actuated for PCV
venting

24

3/16

01:55

3/17

21:00

○

S/C side valve actuation for PCV
venting recorded, but assumed not to25△

PCV venting assumed to have ended at this timing from D/W pressure
increase
7. Operational procedure records in the 5/16 Report

No venting assumed, because no D/W pressure changes confirmed,△

although 7. Records of operational procedures in the 5/16 Report

have been actuatedsays the PCV was being vented at this timing.

S/C side valve closure confirmed for

S/C side valve assumed not to have been closed, because of the D/W

PCV venting△pressure changes, although 7. Operational procedure records in the
5/16 Report says the valve had been closed to end the PCV venting
at 16:05 on 3/15 (No.23).

26

21:30

S/C side valve actuated for PCV
venting

The valve assumed not to have been opened because of the D/W△pressure changes, although 7. Records of operational procedures in
the 5/16 Report says the valve had been actuated.

27

3/18

05:30

S/C side valve closure confirmed for
PCV venting

28

29

3/19

About

S/C side valve actuated for PCV

05:30

venting

11:30

S/C side valve closure confirmed for
PCV venting

30

3/20

About

S/C side valve actuated for PCV

11:25

venting－－－－

Outside the time span of the current analysis, although 7. Operational
procedure records in the 5/16 Report t mentions the PCV venting.
Outside the time span of the current analysis, although 7. Operational
procedure records in the 5/16 Report mentions the PCV venting.
Outside the time span of the current analysis, although 7. Operational
procedure records in the 5/16 Report mentions the PCV venting.
Outside the time span of the current analysis, although 7. Operational
procedure records in the 5/16 Report mentions the PCV venting.

*1 Timings of changing the water injection flow rates and their amounts were set so as not to exceed the daily average flow rates and the total
injection amount, based on the day-to-day records of injection amount to the reactor (Item 7. Compilation of operational procedures in the
May 16 Report).

4.2.

Results of MAAP analysis

4.2.1. Results of Case 1 analysis
Table 10 summarizes the analysis results for Case 1.

Table 10

Summary of Unit-3 analysis results Case 1

Item

Results

Timing of core starting to be uncovered

About 40 hours after the earthquake

Timing of core starting to be damaged

About 42 hours after the earthquake

Timing of RPV being damaged

No RVP damage occurred in this analysis

Details of the Case 1 analysis results follow.
The reactor water levels gradually decreased after the high-pressure coolant
injection (HPCI) had stopped, the core started to be uncovered, the core became
totally uncovered by the SRV actuation, and the core started to be damaged(Figure 15). Water injection started, but the amount of water injected had beenadjusted to simulate the reactor water levels measured; this amount wasinsufficient and was at the level to cover only about half of the core. Consequently,
the core became damaged.
The RPV pressures remained high at around the SRV actuation pressure until
the RCIC and the HPCI stopped. Upon SRV actuation after HPCI shutdown, theRPV pressures dropped quickly and then decreased to the atmospheric pressure(Figure 16).
Concerning the PCV pressures, the D/W pressures and S/C pressurescontinued to increase, because the steam produced in the reactor was discharged
to the S/C. They increased sharply for a while upon SRV actuation, but they
decreased upon the S/C venting. The PCV pressures repeated ups and downs,
thereafter, responding to the venting operations (Figure 17).
The Unit-3 core was concluded to have remained in the core region, althoughpart of the molten fuel remained as a pool, and the RPV was concluded not to havebeen damaged (Figure 18). This was considered to have been feasible because,water injection via the RCIC and HPCI could have been implemented fairly
continuously, and the time delay from the HPCI shutdown to water injection
initiation could have been shorter than at Unit-1.

Attachment 1-32

It should be noted that the RPV pressures showed a decreasing trend while the
HPCI was in service. In the analysis reported in May 2011, steam leaks to outside
the D/W had been assumed via HPCI piping to simulate the RPV pressurechanges and D/W pressure changes. Investigations thereafter, however, showedthat leak paths on the HPCI line were thought to be very unlikely to have formed on
the following grounds: if steam had leaked from the HPCI piping, the R/B would
have been at too high temperatures or in too high temperature steam atmospheres
including the HPCI cell for anybody to access, but actually some operational staff
did access the HPCI cell after the HPCI had stopped on March 13; the HPCI steam
piping was found not to have been damaged by the earthquake from the results ofa seismic evaluation. The RPV pressure changes can be considered to be due tocontinued steam consumption by the HPCI continued operation.

4.2.2. Results of Case 2 analysis
Table 11 summarizes the analysis results of Case 2.

Table 11

Summary of Unit-3 analysis results Case 2

Item

Results

Timing of core starting to be uncovered

About 40 hours after the earthquake

Timing of core starting to be damaged

About 42 hours after the earthquake

Timing of RPV being damaged

About 66 hours after the earthquake

Details of the Case 2 analysis results follow.
The reactor water level gradually decreased after the HPCI had stopped, thecore started to be uncovered, the core became totally uncovered by the SRV
actuation, and the core started to be damaged (Figure 19). Water injection wasstarted, but as an insufficient amount of injected water had been assumed, thereactor water level did not reach above the bottom of active fuel, and this caused
the core damage to be more advanced than in Case 1.
The RPV pressures showed a temporary increase after depressurization by the
SRV actuation. This was due to the steam produced when part of the molten fuelwas relocated to the lower plenum. But the overall behavior (Figure 20) other than
this point was similar to the behavior in Case 1.
The PCV pressures showed a temporary increase, as in the case of RPV

Attachment 1-33

pressures, due to the steam produced when part of the molten fuel was relocatedto the lower plenum. But the overall behavior (Figure 21) other than this point was
similar to the behavior in Case 1.
Part of the molten fuel remained in the RPV, but the RPV was damaged. The
amount of water injected at the beginning had been set at a lower value than that inCase 1 and this caused the core damage to be more advanced (Figure 22).

4.3.

Estimation of Unit-3 core conditions

The core conditions were estimated as follows, by comprehensively considering theinformation available including the core conditions having been estimated from thetemperature behavior actually measured.
The analysis of Case 1 gave the results that the fuel in the Unit-3 core had been molten
but remained in the core region, although part of the molten fuel had remained as a pool,and that the RPV had not been damaged. The analysis of Case 2 gave the results that partof the molten fuel had remained in the RPV, but the RPV had been damaged.
On the other hand, observed plant parameters seemed to have indicated that most of thefuel was being cooled in the RPV based on the following grounds: the RPV steel
temperatures were changing at around 100 to 120 deg C and at more than one
measurement point the temperatures responded in a consistent manner to the changing
amount of water injected; the heat source could be considered to be in the RPV, becausetemperatures at several points showed an increase in May 2011; and the RPV bottom
temperatures changed at around 100 to 170 deg C, the same levels as the temperatures atother locations around the RPV (Figure 23).
Consequently, based on the analysis results and plant parameters, the core wasconsidered to have been significantly damaged and relocated to a location lower than its
original position (lower plenum), and was being cooled there in a stable manner.

Attachment 1-34

10

HPCI started up
RCIC stopped
RCIC停止HPCI起動HPCIHPCI停止
stoppedShroud water level (analysis)
シュラウド内水位(解析）81

Lowered to TAF (about 40 hours after earthquake)
TAF到達(約40時間後)

Downcomer
water level
ダウンカマ水位(解析)(analysis)
Actualwater level indicator

実機計測値

SRV1
opened
SRV1弁開6

readings

BAF到達(約42時間後)
Loweredto BAF (about 42 hours after earthquake)
海水注水停止Seawater
injection stopped

2
TAF

0
-2

BAF

-4
-6

Seawater
injection restarted
海水注水再開Freshwater
injection switched to seawater injection
淡水注入より海水注水に切替-8

淡水注入開始Freshwater injection
started

-10
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

3/16
0:00

3/16
12:00

3/17
0:00

3/17
12:00

3/18
0:00

3/18
12:00

Date
and time
日時

Figure 15

Unit-3 reactor water level changes Case 1

10

HPCI started up
HPCI stopped
RCIC停止 HPCI起動 HPCI停止
RCIC stopped
SRV1弁開SRV1
openedRPV pressureRPV圧力(解析）(analysis)
Actual pressure
indicator readings
実機計測値

8

原子炉圧力 (MPa[abs])

原子炉水位 (m)

4

6

4

2

0

-2
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

3/16
0:00

3/16
12:00

3/17
0:00

Date and
日時timeFigure 16

Unit-3 RPV pressure changes Case 1

Attachment 1-35

3/17
12:00

3/18
0:00

3/18
12:00

SRV1
opened
SRV1弁開S/C vented
S/Cベント1.0

D/W pressures (analysis)
D/W圧力(解析）S/C pressures (analysis)
S/C圧力(解析)Actual pressure
indicator readings (D/W)
実機計測値(D/W)

S/C
ventedS/CベントS/CventedS/Cベント

0.8

S/C vented

原子炉格納容器圧力 (MPa[abs])

S/Cベント計装DS/
Instrumentation
DS hunting
ハンチング
oscillation

S/C
ventedS/Cベント(assumed)
(仮定)

0.6

実機計測値(S/C)
Actual pressure
indicator readings (S/C)

0.4

0.2

0.0

-0.2
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

3/16
0:00

3/16
12:00

3/17
0:00

Date and
日時timeFigure 17

Unit-3 PCV pressure changes Case 1

Attachment 1-36

3/17
12:00

3/18
0:00

3/18
12:00

↑

↑

燃

燃

料

料

部

部

↓

↓

About 64 hours after scram

About 68 hours after scram

↑

↑

燃

燃

料

料

部

部

↓

↓

About 72 hours after scram

About 1 week after scramCore damage schematic: No fuel (collapsed): Intact fuel: Sedimented damaged fuel (pin shape was kept): Fuel pin diameters enlarged by solidified molten fuelwhile going down pin surfaces: Coolant flow paths blocked by enlarged fuel pins
: Molten fuel pool
Figure 18

Unit-3 core conditions Case 1

Attachment 1-37

10

HPCI started up
RCIC停止HPCI起動
RCIC stopped
HPCIHPCI停止
stoppedShroudwater level (analysis)
シュラウド内水位(解析）1

8Downcomer
water level
ダウンカマ水位(解析)(analysis)
Actualwater level indicator
実機計測値
readings

Lowered to TAF (about 40 hours after earthquake)

TAF到達(約40時間後)
6

SRV1 opened
SRV1弁開※：Water levels (analysis) after
Loweredto BAF (about 42 hours after earthquake) RPV damage(about 66
BAF到達(約42時間後)
海水注水停止hours after the earthquake
Seawater
injection stopped(Table 11)) does not mean thewater level is maintained atthat level
TAF

原子炉水位 (m)

4

2

0

-2

BAF
-4

-6

海水注水再開Seawater
injection restarted
-8

-10
3/11
12:00

Freshwater injection switched to seawater injection
淡水注入より海水注水に切替
淡水注入開始Freshwater injection
started
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

3/16
0:00

3/16
12:00

3/17
0:00

3/17
12:00

3/18
0:00

3/18
12:00

Date
and time
日時

Figure 19

Unit-3 reactor water level changes Case 2

10

HPCI started up
HPCI停止HPCIstopped
RCIC停止 HPCI起動
RCIC stopped
SRV1弁開SRV1
opened原子炉圧力 (MPa[abs])

8

RPV pressureRPV圧力(解析）(analysis)
Actual pressure indicator readings

実機計測値

Fuel relocation
to lower plenum (analysis)
下部プレナムへのリロケーション(解析)

6

4

RPV
damaged (about 66 hours after earthquake)
RPV破損(約66時間後)
2

0

-2
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

3/16
0:00

3/16
12:00

3/17
0:00

日時
Date andtime

Figure 20

Unit-3 RPV pressure changes Case 2

Attachment 1-38

3/17
12:00

3/18
0:00

3/18
12:00

SRV1
opened
SRV1弁開S/C vented
S/Cベント
D/W圧力(解析）D/W pressures (analysis)

1.0

S/C圧力(解析)S/C pressures (analysis)
実機計測値(D/W)
Actual pressure indicator readings (D/W)

S/C
ventedS/Cベント0.8

原子炉格納容器圧力 (MPa[abs])

実機計測値(S/C)

Actual pressure indicator readings (S/C)

S/CS/Cベント
ventedS/CventedS/Cベント(assumed)
(仮定)

0.6

S/CS/Cベント
ventedInstrumentation
DS hunting
計装DS/
oscillation
ハンチング

3/15
12:00

3/16
12:00

0.4

0.2

0.0

-0.2
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

3/16
0:00

3/17
0:00

Date andtime
日時

Figure 21

Unit-3 PCV pressure changes Case 2

Attachment 1-39

3/17
12:00

3/18
0:00

3/18
12:00

↑

↑

燃

燃

料

料

部

部

↓

↓

About 58 hours after scram

About 62 hours after scram

↑

↑

燃

燃

料

料

部

部

↓

↓

About 66 hours after scram

About 96 hours after scramCore damage schematic: No fuel (collapsed): Intact fuel: Sedimented damaged fuel (pin shape being kept): Fuel pin diameters enlarged by solidified molten fuelwhile going down pin surfaces: Coolant flow paths blocked by enlarged fuel pins
: Molten fuel pool
Figure 22

Unit-3 core conditions Case 2

Attachment 1-40

1F-3 温度に関するパラメータ（代表点）[Note]

425

Some devices may give incorrect readings in the working
environment
after the earthquake and in the accident progression【留意事項】400

RPV
RPV ｽﾀｯﾄﾞﾎﾞﾙﾄ温度
stud bolt temperature

各計測器については、地震やその後の事象進展の影響を受けて、通常の使用環境条件を超えているも
thereafteras they were operating beyond their normal dutyのもあり、正しく測定されていない可能性のある計測器も存在している。プラントの状況を把握するために、このような計器の不確かさも考慮したうえで、複数の計測器から得られる情報を使用して変conditions.
Readings from multiple devices are interpreted

375

comprehensively in estimating the plant conditions, with their

RPV body flange lower part
RPV 胴ﾌﾗﾝｼﾞ下部温度
temperature

uncertainties taken into consideration.

RPV 胴ﾌﾗﾝｼﾞ
body flange
RPV化の傾向にも着目して総合的に判断している。

350
325

RPVﾍﾞﾛｰｼｰﾙ
RPV seal bellows

300
FeedwaterN4B
nozzleN-4B
給水ﾉｽﾞﾙ
温度

275
Vesselbottom head upper
RPV
底部ﾍｯﾄﾞ上部

250

Main
steam isolation
主蒸気隔離弁
2-86Avalve
ﾘｰｸ
2-86A
ｵﾌ温度

℃

225

S/Cへ

200

Reliefvalve 2-71D
逃し安全弁
2-71D 漏洩

175

逃し安全弁
2-71F 漏洩
Reliefvalve 2-71F

150

圧力容器下部温度
RPV
lower part

125

return line
D/W HVH
HVH戻り温度100

S/C pool temperatures A
S/Cプール水温度A75

S/Cプール水温度B
S/C
pool temperatures B

50
25
0

3/ 20 3/ 23 3/ 26 3/ 29

4/ 1

4/ 4

4/ 7

4/ 10 4/ 13 4/ 16 4/ 19 4/ 22 4/ 25 4/ 28

Figure 23

5/ 1

5/ 4

5/ 7

5/ 10 5/ 13 5/ 16 5/ 19 5/ 22 5/ 25 5/ 28

Unit-3 temperature changes at typical points (May 2011 when disclosed)

Attachment-2

Core conditions estimated from heat balance while water injection was interrupted

The extent of core damage was estimated by comparing the decay heat during the period of no water
injection after the accident, and the amount of coolant which had existed in the RPV at the beginning, or
sensible and latent heat of fuel and in-core structural materials. Table 1 to Table 3 give event sequences
and the time periods when water injection was halted at each unit. Figure 1 shows the decay heat during
the time when water injection was halted, while Figure 2 compares the heat production and heat removal.
Table 4 specifies weights of fuel, structural materials and other items used in the calculation. Decay
heats used at each unit are given in Figure 3 to Figure 5.
At Unit-1, the decay heat released before the seawater injection was started at full capacity far
exceeded the amount which could be absorbed by the water and other materials that had existed in the
RPV. Consequently, the molten fuel that was at elevated temperatures was relocated to the RPV bottom,
where it damaged the RPV and most of the molten fuel would have been relocated to the PCV.
At Unit-2 and Unit-3, the decay heat released while water injection was halted was at the same level
as the amount which could be absorbed by the water and other materials that had existed in the RPV.
Therefore, a certain amount of fuel might have been relocated to the RPV bottom after being molten, but
the RPV would not have been so significantly damaged as to allow a big amount of fuel to be relocatedto the PCV.

Attachment 2-1

Table 1

Operation timeline of water injection to Unit-1 reactor

Day and time
14:46 on March 11, 2011

Operation records
Earthquake occurred
Reactor scrammed

14;47

MSIV closed

14:52

IC(A) (B) started up automatically

15:03

IC(A) stopped
IC(B) stopped

15:17

IC(A) restarted

15:19

IC(A) stopped

15:24

IC(A) restarted

15:26

IC(A) stopped

15:32

IC(A) restarted

15:34

IC(A) stopped

15:37

Station blackout

05:46 on March 12, 2011Freshwater injection started by fire engines

14:55

Freshwater injection terminated

19:04

Seawater injection started

Time of event

Being cooled or water being injected

Time of event

No or little water injection

Attachment 2-2

Table 2 Operation timeline of water injection to Unit-2 reactor
Day and time

Operation records

14:46 on March 11, 2011

Earthquake occurred

14:47

Reactor scrammed

15:02

Reactor Isolation Cooling System (RCIC) manually started up

15:28

RCIC tripped (L-4)

15:41

Station blackout

13:25 on March 14, 2011

RCIC stopped

16:34

RPV depressurization started (ARV1 valve opening)
Seawater injection started via Fire Protection System

About 18:00

RPV pressure decrease confirmed

19:20

Fire engines stopped due to fuel shortage

19:54

First fire engine started pumping again

19:57

Second fire engine started pumping again

21:20

RPV depressurized by SRV2 opening, water level recovered

Time of event

Being cooled or water being injected

Time of event

No or little water injection

Attachment 2-3

Table 3 Operation timeline of water injection to Unit-3 reactor
Day and time

Operation records

14:46 on March 11, 2011

Earthquake occurred

14:47

Reactor scrammed

15:06

RCIC manually started up

15:25

RCIC tripped (L-8)

15:38

Station blackout

16:03

RCIC manually started up

11:36 on March 12, 2011
12:35
02:42 on March 13, 2011
About 09:08

RCIC tripped
HPCI started up (L-2)
HPCI stopped
RPV depressurization operation by SRVs

09:20

PCV pressure decrease due to PCV venting confirmed

09:25

Freshwater injection started

Time of event

Being cooled or water being injected

Time of event

No or little water injection

Table 4 Weights used in the calculation (structures, fuel, water inventory)
Unit-1

Unit-2, Unit-3

CR guide tube

12.3

t

17.4

t

CRD housing (in the core)

4.6

t

6.5

t

In-core probe guide tube

0.8

t

1.1

t

In-core probe housing

0.2

t

0.3

t

In-core probe stabilizer (in the core)

0.1

t

0.1

t

18

t

25

t

UO2

79

t

107

t

Zircaloy

32

t

43

t

Water below bottom of fuel

52.4

t

72.6

t

Water between fuel bottom and normal level)

93

t

145

t

Injected water

80

t

0

t

Total core internals (Sum of the above) *(Water inventory)* Rounded off to the first number after the decimal point

Attachment 2-4図１．崩壊熱（ORIGENコードによる）

30

Note: At Unit-1, about 80 tons of water25

injected from early morning to daylight on
March 12, 2011. This was added to theoriginal inventory for comparison.

15

10

Unit-3

３
号
機

5

２
号
機

Unit-2

１
号
機
Unit-1

Decay heat (MW)
崩壊熱（MW）

20

0

Figure 1

図２． 崩壊熱量と除熱量の比較
図２． 崩壊熱量と除熱量の比較
1000

800

800

400

崩壊熱量(GJ)

1000

600

Decay heat changes and timings of no water injection at each unit

Zircaloy(sensible +latent heat)
ジルカロイ（顕熱+融解熱）
ジルカロイ（顕熱+融解熱）
UO2（顕熱+融解熱）
UO2UO2（顕熱+融解熱）(sensible +latent heat)
構造材（顕熱+融解熱）
構造材（顕熱+融解熱）
Structures(sensible +latent heat)
注水
注水
Injected
water
崩壊熱
崩壊熱
炉水（燃料下端以下）
Decay
heat
炉水（燃料下端以下）
炉水（燃料下端以上）
Reactorwater (below BAF)
炉水（燃料下端以上）

600

Reactor water (above BAF)

400

200

200
0
1

１号機
2
Unit-1

3

4

4

２号機
5

２号機
5
Unit-2

6

7

３号機
8
Unit-3

9

0
1

１号機
2
3
Figure 2

6

7

３号機
8

Comparison of decay heat and heat removal capacities

Attachment 2-5

9

30

25

Decay heat (MW)
崩壊熱（MW）

20

15

10

5

0
3/11 0:00

3/11 12:00

3/12 0:00

3/12 12:00

Figure 3

3/13 0:00

3/13 12:00

3/14 0:00

3/14 12:00

3/15 0:00

3/14 0:00

3/14 12:00

3/15 0:00

Unit-1 decay heat

30

25

Decay heat (MW)
崩壊熱（MW）

20

15

10

5

0
3/11 0:00

3/11 12:00

3/12 0:00

3/12 12:00

Figure 4

3/13 0:00

3/13 12:00

Unit-2 decay heat

Attachment 2-6

30

Decay heat (MW)
崩壊熱（MW）

25

20

15

10

5

0
3/11 0:00

3/11 12:00

3/12 0:00

3/12 12:00

Figure 5

3/13 0:00

3/13 12:00

Unit-3 decay heat

Attachment 2-7

3/14 0:00

3/14 12:00

3/15 0:00

This document includes some undisclosed
data, not contained in the “Plant_data” of
the TEPCO homepage.

Attachment-3

Core conditions estimated from the temperatures and pressures measured
1.

Outline of reactor cooling conditions and temperature and pressure behaviors
At each unit, water injection started after the accident using the accident management(AM) lines and other paths, the drywell (D/W) pressure measurements started immediately
after the accident, and the temperature measurements started from the second half ofMarch 2011. Cooling conditions, and temperature and pressure behaviors at each unit aresummarized below.
(1) Unit-1
At Unit-1, water injection started on the day after the earthquake (March 12, 2011) via
the fire protection systems, followed by the injection via the feedwater system (FWS),
which had been placed in service from March 23, 2011. Nitrogen injection to the D/Wstarted on April 7, 2011.
In May 2011, the amount of water injection was temporarily increased to test for the
possibility of flooding the containment vessel (PCV) with water. Thermometer readings
responded roughly consistently to the changing amount of water being injected. The D/Wpressure indicator readings responded clearly to the amount of water injected, i.e., thetendency of pressure decrease with increased water injection and pressure increase with
reduced water injection was clearly recognized.
Thereafter, the amount of water injection was reduced, in response to the decreasing
decay heat. This was necessary to cope with the concern that the amount ofcontaminated water retained on the site was increasing, and the amount of water injection
was being optimized (3.5 to 4.0m3/h was being maintained).
Since August 2011, water injection of 3.5 to 4.0m3/h has been continued, and theoverall temperatures, including those of the reactor vessel (RPV) and PCV cut below 100
deg C and have continued to decrease gradually.
In late October 2011, the water injection rate via the FWS was increased to 7.5m3/h toensure control of steam generation. The RPV and PCV temperatures decreased further
accordingly.
The water injection rare was temporarily decreased to 5.5m3/h in preparation for placing
the core spray (CS) system in service. The RPV and PCV temperatures are currently (asof November 21, 2011) changing around 40 deg C.
Figure 1 shows the temperature trends at typical points of the RPV and PCV, while
Figure 2 shows the D/W pressure trends.

Attachment 3－1

Water injection increased

CS water injection preparation

Figure 1

Unit-1 RPV and PCV temperature trends(A: RPV flange, B: Seal bellows, C: Feedwater nozzle, D: RPV lower part, E: D/W-HVH)

Nitrogen gas injection

Figure 2

Unit-1 D/W pressure trends(2) Unit-2
At Unit-2, water injection started via the fire protection system 2 days after theearthquake. In late May 2011, the FWS was placed in service for water injection. The
water injection continued by the FWS after switching from the fire protection system. Afterthe FWS was placed in service for water injection, temperatures of the RPV bottom (upper

Attachment 3－2

face of the RPV bottom head) were changing around the saturation temperature, but at
the RPV upper and middle parts temperature behavior indicated superheated conditions.
Thereafter, the amount of water injection was reduced, in response to the decreasing
decay heat. This was necessary to cope with the concern that the amount ofcontaminated water retained on the site was increasing, and the amount of water injection
was being optimized (3.5 to 4.0m3/h was being maintained). On June 28, 2011, nitrogengas injection started.
From the middle of September 2011, the CS system started water injection in parallelwith the FWS, and this was intended to improve water injection efficiency. The water
injection path from the CS system goes directly through the core region. The D/Wpressures increased temporarily due to the water injection from the CS system. This isconsidered as being caused by the water injected from the CS system contacting thesuperheated structures in the RPV and mostly generating steam within a limited time.
Thereafter, the D/W pressures began to decrease with the increasing amount of water
being injected. In addition, the degree of superheating at the RPV upper part began todecrease as soon as the CS system had been placed in service for water injection.
In late September 2011, the amount of CS system water injection was increased in
response to the improved service conditions of the contaminated water processing
system. Since early October 2011, almost all RPV and PCV thermometers have been
indicating lower temperatures than the D/W saturation temperature, with the exception ofsome thermometers outside the RPV indicating higher temperatures locally.
It is noted that the D/W pressures began a gradual increase in late September 2011.
This is estimated to be due to nitrogen injection, when the overall temperature behavior isconsidered.
Figure 3 shows the temperature trends after March 2011 at some typical points of the
RPV and PCV, while Figure 4 shows the D/W pressure trends.

Attachment 3－3

CS water injection

Figure 3

Unit-2 RPV and PCV temperature trends(A: RPV flange, B: Seal bellows, C: Feedwater nozzle, D: RPV lower part, E: D/W-HVH)

Nitrogen gas injection

PCV gas control system in service
ガス管理システム起動

Figure 4

Unit-2 D/W pressure trends(3) Unit-3
At Unit-3, water injection started via the fire protection system 2 days after theearthquake. In the middle of May 2011, the FWS was placed in service for water injection.
The water injection continued by the FWS after being switched from the fire protection
system. After the FWS was placed in service for water injection, temperatures at various

Attachment 3－4

points of the RPV continued to indicate superheated conditions. The temperatures havebeen changing in a stable manner since June 2011 at a higher level.
Thereafter, the amount of water injection was reduced, in response to the decreasing
decay heat. This was necessary to cope with the concern that the amount ofcontaminated water retained on the site was increasing, and the amount of water injection
was being optimized (reduced from 13.5 m3/h to 4.0 m3/h with time). On July 14, 2011,
nitrogen gas injection was started.
In early September 2011, the CS system started water injection in parallel with the FWS,
and this was intended to improve water injection efficiency. The water injection path fromthe CS system goes directly through the core region. The degree of superheating over thewhole RPV began to decrease as soon as the CS system had been placed in service forwater injection.
In the middle of September 2011, the amount of CS system water injection wasincreased in response to the improved service conditions of the contaminated water
processing system. Consequently, in late September 2011 most of the RPV
thermometers including the one at the reactor bottom (bottom head) gave readings below
100 deg C. Until early October 2011, some thermometers outside the RPV indicated hightemperatures locally, but now (November 2011) thermometers at all measurement points
indicate readings below 100 deg C.
It is noted that the D/W pressures were changing at the level around the atmosphericpressure since March 2011, but that no change has been noticed although nitrogen gas
injection started in July 2011 (the pressure indicator was switched over at the same time).
Figure 5 shows the temperature trends after March 2011 at some typical points of the
RPV and PCV, while Figure 6 shows the D/W pressure trends.

Attachment 3－5

CS water injection

Figure 5

Unit-3 RPV and PCV temperature trends(A: RPV flange, B: Seal bellows, C: Feedwater nozzle, D: RPV lower part, E: D/W-HVH)

Nitrogen gas injection

Figure 62.

Unit-3 D/W pressure trends

Analysis from temperature behavior
Cooling conditions discussed in the previous section and temperature and pressure

behaviors have been analyzed. The results are summarized below.

(1) Unit-1[Analysis of temperature behavior from March to May 2011]

Attachment 3－6

When the capability to measure RPV temperatures was recovered in late March, theRPV temperatures were above 400 deg C at more than one point. At that time, the core isconsidered not to have been sufficiently cooled. But as soon as the water injection path to
the reactor was changed to the FWS, the water was assured to reach the reactor andtemperatures at various points decreased rapidly. The core was then considered to be
being sufficiently cooled.
On the other hand, most of the fuel was considered to have been cooled in the RPV
based on the following reasons: temperatures of CRD housings and other structures
below the RPV were measurable, which would not be possible if the RPV had beendamaged; the RPV steel temperatures were changing at the level around 100 to 120 deg
and the temperatures at several points responded consistently to the changes of theamount of injected water; and the temperatures at the RPV upper part were higher atseveral points indicating the presence of the heat source in the RPV (Figure 7).
In conclusion, the core was considered, from plant parameters (temperatures aroundthe RPV), to have been relocated from its original position to below (lower plenum) andmostly being cooled there in a stable manner, even if the core had been largely damaged.
It should be noted, however, that the temperature changes over an extended time tillnow (November 2011) indicated the rapid decrease of high temperatures observed at the
RPV upper part in March and April 2011 being similar to the attenuation behavior of shorthalf-life nuclides such as iodine-131 and others. This may mean that the hightemperatures at the RPV upper part might have been caused by volatile radioactive
materials that had deposited.

Figure 7

CRD housing temperature trends

Attachment 3－7

[Analysis of temperature behavior from May to October 2011]
Temperatures of CRD housings and other structures below the RPV could be measured
after May 2011, too. Measured temperatures at several points responded to the changes
in the amount of injected water, as had been existing until May 2011. The following
findings were newly obtained.


The temperature difference between the RPV upper part and bottom decreased.



The RPV bottom temperatures decreased to below the saturation temperature andhave continued to decrease as of August 2011.
The fuel was considered not to be present in the core region, because the reduced

temperature difference at the RPV upper and lower parts indicated that there is little
uncovered fuel in the RPV, and also by considering the injected water path, which is notgoing through the core region, and the calibrated water level indicator readings. It should
be noted in addition that the amount of water being injected was not sufficient to remove
sensible heat, nevertheless the RPV bottom temperatures cut below the saturation
temperature. Most of the fuel was considered not to be present in the RPV.

[Analysis of temperature behavior after October 2011]
The following findings were newly obtained by increasing the amount of water injection
in late October.


RPV and PCV temperatures rapidly decreased by cooling. Meanwhile, thesuppression chamber (S/C) pool temperatures increased after the amount of injectedwater increased and the temperatures of the RPV and PCV and S/C pool werereversed (Figure 8).
If water flowed into the S/C from the RPV after heat exchange only in the RPV, the S/Ctemperatures should not exceed the RPV bottom temperatures, but actually the S/Ctemperatures exceeded the RPV bottom temperatures. This may indicate the possibility
that a heat source existed in the PCV lower part (probably around the pedestal) and thatthe water flowed into the S/C through the vent tube after having contacted the heat sourcein the PCV. The S/C pool temperature increase is considered to have been caused bymore hot water inflows to the S/C, when the water injection amount had been increased,
carrying more energy equivalent to the decreased energy necessary for steam generationin the PCV.

Attachment 3－8

Figure 8

Relationships between RPV and PCV temperatures and S/C pool temperatures(2) Unit-2[Analysis of temperature behavior from March to May 2011]
Most of the fuel was considered to be being cooled in the RPV based on the following
reasons: the RPV bottom temperatures were changing in the range of about 100 to 120
deg C and responded consistently at multiple points to the changing amount of injectedwater; and the temperatures at the RPV upper part were higher indicating the presence ofa heat source in the RPV.
Consequently, the core was considered, from plant parameters, to have been relocatedfrom its original position to below (lower plenum) and mostly being cooled there in a stablemanner, even if the core had been largely damaged.

[Analysis of temperature behavior from May to September 2011]
The following findings were newly obtained from the RPV and PCV temperature
behavior since May 2011.
The RPV bottom temperatures indicated roughly readings around the saturation
temperature since the water injection via the FWS started in May 2011, but the RPV
upper and middle parts remained at higher temperatures.
From this observation, part of the uncovered fuel was considered to be present at thecore region, overheating the RPV inside, although the cooling effect by the water from the
FWS was confirmed.

[Analysis of temperature behavior from September 2011]
The following findings were newly obtained from the RPV and PCV temperature
behavior since September 2011.


Temperatures at the RPV upper part decreased due to the water directly passing

Attachment 3－9

through the core region from the CS system, and further decreased to below
saturation temperature when the amount of water injection had been increased.


Temperatures of the PCV atmosphere remain mostly below the saturation
temperature, but very locally (CRD housings, SRVs and others) some thermometers
still indicate high temperatures (above the saturation temperature).

From these observations, the following interpretations may be possible.
Part of the fuel is present in the RPV core region, but most of the fuel is under waterin the RPV lower part.

Heat sources are present outside the RPV, too, but they are being cooled sufficiently.
However, some fuel may be uncovered (around CRD housings) and some parts may
be generating mild heat from the deposited volatile fission products and otherspecies (around the SRVs) (Figure 9 and Figure 10).

Figure 9

CRD housing temperature trends
Safety valve (A) Leak detector

Safety valve (B) Leak detector

Safety valve (C) Leak detector

SRV (A) Leak detector

SRV (B) Leak detector

SRV (C) Leak detector

SRV (D) Leak detector

SRV (E) Leak detector

SRV (F) Leak detector

SRV (G) Leak detector

SRV (H) Leak detector

Figure 10

Safety valve and SRV leak detector temperature trends

Attachment 3－10

(3) Unit-3[Analysis of temperature behavior from March to May 2011]
Most of the fuel was considered to be being cooled in the RPV based on the following
reasons: the RPV steel temperatures were changing in the range of about 100 to 200 deg
C and responded consistently at multiple points to the changing amount of injected water;
the heat source was estimated to be present in the RPV, because temperatures increased
at multiple points in May 2011; and the temperatures at the RPV bottom were changing in
the range of about 100 to 170 deg C, similar to the temperatures at other points aroundthe RPV.
Consequently, the core was considered, from plant parameters, to have been largely
damaged and relocated from its original position to below (lower plenum) and mostly
being cooled there in a stable manner.

[Analysis of temperature behavior from May to August 2011]
The following findings were newly obtained from the RPV and PCV temperature
behavior since May 2011.
Overheated conditions of the RPV as a whole continued, although the water injection
by FWS had been placed in service.

The S/C temperatures increased when the amount of water injection was increased(Figure 11).
From these observations, part of the uncovered fuel was considered to be present in

the core region, overheating the RPV inside, because no significant cooling effect was
confirmed in the upper region of the core by the water from the FWS and someoverheated parts remained in the RPV upper region of the RPV. Part of the fuel was
considered to be present in the lower plenum as well as on the injected water path fromthe FWS, because the S/C temperatures increased.

[Analysis of temperature behavior from September 2011]
The following findings were newly obtained from the RPV and PCV temperature
behavior since September 2011.


Temperatures at the RPV upper part decreased due to the water directly passing
through the core region from the CS, and further decreased to below 100 deg C inlate September by increasing the amount of water injection

Temperatures of the PCV atmosphere remained mostly below the saturation
temperature, but for a while very locally (RPV seal bellows and SRVs) somethermometers still indicated high temperatures (above the saturation temperature)

Attachment 3－11

even after the RPV temperatures were lowered below 100 deg C (Figure 12 andFigure 13).
From these observations, the following interpretations may be possible.
Part of the fuel is present in the RPV core region, but most of the fuel is under waterin the RPV lower part.

Heat sources are present outside the RPV, too, but being cooled sufficiently.
However, some fuel may be uncovered (around CRD housings) and some parts may
be generating mild heat due to the deposited fission products (around the SRVs).

S/C gas 114T
S/C gas 114U
S/C gas 114V
S/C gas 114W
S/C temperature ESS-Ⅰ
S/C temperature ESS-Ⅱ

Figure 11

S/C temperature trends

Seal bellows temperatures L
Seal bellows temperatures M
Seal bellows temperatures N
Seal bellows temperatures P

Figure 12

RPV seal bellows temperature trends

Attachment 3－12

Figure 13

3.

SRV leak detector temperature trends

Fuel locations estimated from temperature behavior
Summarized below are the fuel locations at each unit estimated from the analyses in thisdocument.
(1) Unit-1
As of May 2011, most of the fuel was considered to have been being cooled in the RPVfrom the following reasons: temperatures of CRD housings and other structures below theRPV bottom were being measured; RPV steel temperatures were changing at hightemperatures and responded at multiple points consistently to the changing amount ofwater injection; and heat sources were estimated to be present in the RPV, becausetemperatures at the RPV upper part were high at multiple points.
Therefore, the core was considered to have been relocated from its original location to
below (lower plenum) and being cooled there in a stable manner, even if the core had
been largely damaged.
However, the RPV bottom temperatures decreased to below the saturation temperaturefrom May 2011, notwithstanding that the amount of water injection had been insufficient toremove sensible heat. Then part of the fuel was considered to be very likely not present in
the RPV.
Furthermore, in October 2011, when the amount of water injection was increased, theS/C pool temperatures increased and exceeded the RPV and PCV temperatures.
Presently (November 2011), the heat source is considered to be present in the lower partof the PCV (probably in the pedestal area).

Attachment 3－13

(2) Unit-2
As of May 2011, most of the fuel was considered to have been being cooled in the RPVfrom the following reasons: temperatures of the RPV bottom were changing at the level ofabout 100 to 120 deg C and they responded at multiple points consistently to the changing
amount of water injection; and the heat source was estimated to be present in the RPV,
because temperatures at the RPV upper part were higher. Consequently, the core wasconsidered to have been relocated from its original position to below (lower plenum) andbeing cooled there in a stable manner, even if the core had been largely damaged.
The RPV and PCV temperature behavior since September 2011 shows that thetemperatures at the RPV upper part decreased due to the water injected by the CS system
directly running through the core region and they cut below the saturation temperature,when the amount of water injection was increased. This also supports the estimation thatmost of the fuel is under water in the RPV lower part, although a limited amount of fuel ispresent at the RPV core region.
It should be noted that some heat source present outside the RPV is considered as being
sufficiently cooled.

(3) Unit-3
As of May 2011, most fuel was considered to have been being cooled in the RPV fromthe following reasons: temperatures of the RPV bottom were changing at the level ofabout 100 to 200 deg C and they responded at multiple points consistently to thechanging amount of water injection; and the heat source was estimated to be present in
the RPV, because temperatures at the RPV upper part were higher. Consequently, thecore was considered to have been relocated from its original position to below (lower
plenum) and being cooled there in a stable manner, even if the core had been largely
damaged.
The RPV and PCV temperature behavior since September 2011 shows that thetemperatures at the RPV upper part decreased due to the water injected by the CS
system directly running through the core region and cut below the saturation temperaturewhen the amount of water injection was increased. This also supports the estimation thatmost fuel is under water in the RPV lower part, although a limited amount of fuel is present
at the RPV core region.
It should be noted that some heat source present outside the RPV is considered as
being sufficiently cooled.
End

Attachment 3－14

Attachment-4

Core fuel temperatures estimated by the in-RPV temperature evaluation model
1.

Introduction
Temperatures of the core and structural materials in the reactor vessel (RPV) were

estimated by evaluating the heat balance using an in-RPV temperature evaluation model.
Uncertainties in the input data were considered by changing the data parametrically in acertain range and the measured values obtained to date were reviewed for evaluation.
This evaluation model is based on an energy transfer model in which the heat energy istransferred to each structural component by the steam produced. A case with little steam
production is outside the scope of applicability. For this reason, Unit-1 was excluded fromthe evaluation, because the temperatures around the Unit-1 RPV were low. In-RPV
temperatures of Unit-2 and Unit-3 were evaluated.

2.

Temperature evaluation model in the RPV

2.1.

Outline of the evaluation model

Figure 1 shows the configuration used in the model. Heat is generated by the fuel in theflooded and uncovered (unflooded) regions and transferred to the upper structures, coreshroud, RPV upper wall and side walls. In the figure, coolant flows from the water injectionlines are shown in solid lines, while broken lines show the heat flow from the fuel in theuncovered region.
In the model, heat is generated in the fuel in the RPV lower plenum (flooded), the fuel inthe original core region (uncovered) and the fuel that had been relocated to the containment
vessel (PCV). But the heat from the fuel in the PCV is assumed not to contribute to thesteam production in the RPV. The fuel is cooled by the water injected from two systems, i.e.,
the feedwater system and the core spray (CS) system.
Water injected from the feedwater system goes through the lower plenum and becomessaturated steam while being heated in the flooded region, and further, becomessuperheated steam at fuel surface temperatures by being heated in the uncovered region.
On the other hand, water injected from the CS system (CS water) is considered to condense
the superheated steam in the upper core region. In the model, the heat to be removed forcondensation is considered as negative heat generation in the uncovered core region. Partof the water from the CS system becomes steam by being heated up in the uncovered coreregion and reaches the lower plenum, after transferring energy (heat) to the core shroud andupper structures, for steam production in the flooded region.

Attachment 4-1

The uncovered core is cooled by the heat transfer to the saturated steam produced in theflooded core region and by the heat radiation transfer to the core shroud and upper
structures. The heat transferred to the superheated steam and upper structures is furthertransferred to the RPV upper wall by natural convection heat transfer of single-phase steam
or by heat radiation, and then removed to the drywell (D/W) by natural convection heattransfer via heat conduction in the RPV upper wall. The heat transferred to the core shroud
is removed to the D/W by natural convection heat transfer via heat conduction in the coreshroud, heat radiation transfer from the core shroud to the RPV side walls and heat
conduction in the RPV side walls.
By solving the heat balance equation based on the scenario above, temperatures of theuncovered core and various points in the RPV can be calculated.

Attachment 4-2

Tamb
Qncuh
RPV
upper head
RPV上部壁

Quh

QraduhUpper internal
structures
上部構造材

Qfuh

Qfs

Mfed
Tin
Water
injected via
給水系から
feedwaterの注水system

Qs
Water
injected via CS
CS系からの注水

Mcs
Tin

Qradu

Mgc

Condensation
CS水によるby CS凝縮
water-Δ Qcsα

Tamb

Qfsh

Qsw Qradsw

Uncovered

Qsh

Qncsw

QradshRPVsidewall
RPV側壁

炉
心
シ
ュ
ラ
ウ
ド

露出炉心
Qc
core

Qfc

Mgc,csα
Mglp

Qlp
Flooded fuel冠水した燃料

Mfed+(Mcs-Mgc,csα )
-Mglp

Heat
paths
熱の流れ
Water
paths
水の流れ

See the text for図中の変数名は本文参照
variables in the figure.

Figure 1

Schematic of in-RPV temperature evaluation model

Attachment 4-3

2.2.

Methods used for the temperature evaluation model in the RPV

2.2.1. Heat balance equations
The following heat balance equations are used in the evaluation.
Heat balance equation in the uncovered coreQc  Qradsh  Qradu  Q fc’

Qc

Heat generated in the uncovered core (CS water considered)

QradshHeat radiation from the uncovered core to core shroud wall

Qradu

Heat radiation from the uncovered core to the lower surface of(Eq.1)

upper structures
Qfc

Heat received by steam passing through the uncovered core

Heat balance equation from the uncovered core to the RPV upper wall

Qncuh  Quh  Qraduh  Q fuh(Eq.2)

Qraduh  Qs  Qradu  Q fs

(Eq.3)

Qncuh

Heat transfer from the RPV outer surface to D/W atmosphere

QuhHeat conduction from the RPV upper wall inner surface to its
outer surface

QraduhHead radiation from upper structures to the RPV upper wall innersurface

Qfuh

Heat transfer from steam running through upper structures to the
RPV upper wall inner surface

Qs

Heat conduction from the lower surface of upper structures totheir upper surface

Qfs

Heat transfer from steam running through the uncovered core toupper structures

Heat balance equation from the uncovered core to RPV side walls

Qncsw  Qsw  Qradsw  Qsh  Qradsh  Q fsh
Qncsw

Heat transfer from the RPV side wall outer surface to D/W
atmosphere

Qsw

Heat transfer from the RPV side wall inner surface to its outersurface

Qradsw

Heat radiation from the core shroud wall outer surface to the RPVside wall inner surface

Qsh

Heat transfer from the core shroud wall inner surface to its outersurface

Attachment 4-4(Eq.4)Qfsh

Heat transfer from steam in the uncovered core to the coreshroud wall

2.2.2. Amount of steam production and heat generated in the uncovered core region
In the current model, the heat is generated in the fuel in the RPV lower plenum (flooded),
the fuel in the original core region (uncovered) and the fuel that had been relocated to the
PCV. The heat from the fuel in the PCV is assumed not to contribute to the steam productionin the RPV. Heat generation in each region is given by the following equations.

Qd  Qc  Qlp  Qpcv

(Eq.5)

Qlp  X f Qd

(Eq.6)

Qpcv  X s Qd

(Eq.7)

Qc  1  X f  X s Qd
Qd

Fuel decay heat

Qc

Heat generation of uncovered fuel

Qlp

Heat generation of flooded fuelQpcv

Heat generation of relocated fuel to the PCV

XfFraction of fuel in the flooded region

Xs

Fraction of fuel that had been relocated to the PCV (Fraction of(Eq.8)

relocated fuel to the PCV)

Water injected from the feedwater system and CS system is mixed, warmed andevaporated as steam in the flooded region in the lower plenum. The amount of steam
produced in the lower plenum is given by the next equation.

M glp Qlp  1   wl M fed C pw Tsat  Tin   Qlcs 
h fg

Mglp

Steam production in the lower plenum

wlFraction of injected water not contributing to cooling (Fraction ofwater leak)

Mfed

Amount of water injected from the feedwater system

Cpw

Specific heat of water

Tsat

Saturation temperature

Tin

Injected water temperatures

ΔQlcsαHeat energy needed for CS water that had flowed into the lower
plenum to become saturated water

Δhfg

Latent heat

Attachment 4-5(Eq.9)

The amount of steam passing through the uncovered core is the sum of the amount ofsteam produced in the lower plenum and that produced in the uncovered core region fromthe water injected from the CS system.

M gc  M glp  M gc,cs

(Eq.10)

Mgc

Amount of steam passing through the uncovered fuel region

Mgc,csαAmount of CS water evaporation in the uncovered fuel region

Concerning the heat generation in the uncovered core fuel, part of the water injected fromthe CS system is considered to condense the superheated steam and that heat forcondensation is, in the current model, regarded as negative heat generation in theuncovered fuel. The logical background of this model is the following.
’

The amount of heat generated by the uncovered fuel Qc is given by the following
equation. When Q1 is larger than Qc, the whole uncovered fuel is being cooled by the water
injected from the CS system.

0
Qc  
Qc  Q1

Q1  Qc 
Q1  Qc 

Q1  M cs h fg  C pw Tsat  Tin 
Q1(Eq.11)(Eq.12)Heat energy needed for CS water contributing to cooling the
unflooded fuel in the core region to become saturated steamΑ

Direct cooling rate of uncovered core by CS water (Fraction ofeffective CS contribution)

Mcs

Amount of water injected from CS

Part of the CS injection water flows down to the lower plenum, after contributing to thecondensation of the superheated steam. Its temperature is given by the following equation.
It means that the CS water flows down to the lower plenum as saturated water if the heat
energy Q2 necessary for the CS water to reach the saturation temperature is less than theenergy received by the CS water from the uncovered core, otherwise it exists as sub-cooled
water.

 Tsat
Tlcs  T  Qcs
in
M cs C pw


Qc  Qc  Q2 
Qc  Qc  Q2 

Q2  M cs C pw Tsat  Tin 
Tlcs

Temperature of CS water flowing to the lower plenum

Attachment 4-6(Eq.13)(Eq.14)

Heat energy needed for CS water to become saturated water

Q2

The amount of steam produced by heat removal from the uncovered core by the CS water
can be expressed in the following form.

M gc,cs Qc  Qc   M cs C pw Tlcs  Tin h fg(Eq.15)The amount of CS water flowing down to the lower plenum is the sum of the water which
has contributed to the cooling of the uncovered core (minus the amount evaporated as
steam) and the water which has not contributed. The heat energy necessary for this water(the sum) to become saturated water is given by the next expression.

Qlcs  M cs  M gc,cs C pw Tsat  Tlcs   1   M cs C pw Tsat  Tin (Eq.16)2.2.3. Heat transfer from the uncovered core to RPV upper wall
The heat generated in the uncovered core and the heat transferred to the steam QR
passing through the core have the following relationship.

Q fc  M gcC pg Tc  Tsat (Eq.17)

Cpg

Specific heat of steam

Tc

Steam temperatures heated up in the uncovered core (fuelsurface temperatures in the uncovered core)

Heat transfer from the superheated steam heated in the uncovered core to the upper
structures is expressed in the following equation.

Q fs  M gcC pg Tco  Tst 
Tco(Eq.18)

Steam temperature in the uncovered core with consideration ofheat transfer to the core shroud walls

Tst

Temperature on the upper surface of upper structures

The amount of heat radiation transfer Qradu from the uncovered core to upper structures
and the amount of heat conduction Qs from the lower surface to upper surface of upper
structures are expressed in the following equations.

Qradu  Act

1

c



Qs  As s

1
1

s

1

 Tc4  Tsb4 

Tsb  Tsts

Attachment 4-7(Eq.19)(Eq.20)

Act

Upward projection area of the uncovered coreεcEmissivity of the coreεs

Emissivity of upper structuresσStefan-Boltzmann constantTsb

Temperature of upper structure lower surface

As

Heat transfer area of upper structuressHeat conduction coefficient of upper structures

Tst

Temperature of upper structure upper surface

δsThickness of upper structures

In evaluating the heat transfer from the upper structures to the RPV upper wall, heatradiation from the upper structures and heat transfer by the steam passing through theupper structures are considered. The amount of heat transfer by heat radiation Qraduh can be
expressed by the next equation.

Qraduh  As

1

s



1
1

 uh

1

4

 Tst4  TuhinεuhEmissivity of RPV upper wall

Tuhin

Temperature of RPV upper wall inner surface(Eq.21)

Steam temperature Tuh,g passing through upper structures can be obtained from thefollowing two heat balance equations Qfuh. One is the heat balance between the steam and
RPV upper wall, and the other is the heat balance when the steam passes through theupper structures.

Q fuh  AuhhuhinTuh, g  Tuhin 

Q fuh  M gcC pg Tst  Tuh, g 

AuhHeat transfer area of RPV upper wall

huhin

Heat transfer coefficient of RPV upper wall

(Eq.22)(Eq.23)

Concerning the heat transfer from the RPV to the D/W atmosphere, heat conduction Quhfrom the RPV upper wall inner surface to outer surface, and heat transfer Qnuch from the
RPV upper wall outer surface to D/W atmosphere are considered.

Quh  AuhuhTuhin  Tuho

 uh
Qncuh  Auhhnc Tuho  Tamb uhHeat conduction coefficient of RPV upper wall

Attachment 4-8(Eq.24)(Eq.25)TuhoTemperatures of RPV upper wall outer surface

δuho

Thickness of RPV upper wall

hnc

Heat transfer coefficient of natural convection

Tamb

D/W atmosphere temperatures

2.2.4. Heat transfer from the uncovered core to RPV side walls
Heat transfer from the uncovered core to the in-RPV steam and to the core shroud innerwall is obtained from the following heat balance equations. One is the heat balance between
the in-RPV steam and core shroud inner wall, and the other is the heat balance when thein-RPV steam is cooled after having passed through the uncovered core.

Q fsh  Ashhshin Tco  Tshin 
Q fsh  M gcC pg Tc  Tco 

AshHeat transfer area of the core shroud wall

hshin

Heat transfer coefficient of the core shroud wall

Tco

In-RPV steam temperature

Tshin

Temperature of the core shroud wall inner surface(Eq.26)(Eq.27)

The amount of heat transfer Qradsh by heat radiation from the uncovered core to the coreshroud inner wall is given by the next equation.

Qradsh  Acs

1

c



1
1

 sh

1

4

 Tc4  Tshin

Acs

Side wall area of the uncovered coreεshEmissivity of the core shroud wall

(Eq.28)

Four components are considered for the heat transfer from the core shroud inner wall to
the D/W atmosphere: heat conduction from the core shroud inner surface to the outersurface Qsh; heat radiation from the core shroud outer surface to the RPV side wall innersurface Qradsw; heat conduction from the RPV inner surface to the outer surface Qsw; andheat transfer from the RPV side wall outer surface to the D/W atmosphere Qncsw.

Qsh  Ashsh
Qradsw  Ash

1

 sh



Tshin  Tsho

1
1

 sw

 sh

1

4
4

 Tsho TswinAttachment 4-9(Eq.29)(Eq.30)Qsw  Aswsw

Tswin  Tswo sw
Qncsw  Aswhnc Tswo  Tamb 

3.

sh

Heat conduction coefficient of the core shroud wall

Tsho

Temperature of the core shroud wall outer surface

δshoThickness of the core shroud wall

εsw

Emissivity of the RPV side wall

Tswin

Temperature of the RPV side wall inner surface

Asw

Heat transfer area of the RPV side wallsw

Heat conduction coefficient of the RPV side wall

Tswo

Temperature of the RPV side wall outer surfaceδsw

Thickness of the RPV side wall

(Eq.31)(Eq.32)Estimation of core fuel debris conditions
The in-RPV temperature evaluation model developed above (2.2) was applied toreproduce measured data to date, by which the debris distribution in the RPV was estimated.
The estimated results were then used to evaluate the in-RPV temperature distributions as of
October 10, 2011.

3.1.

Selection of time points (date and time) and temperatures for evaluation

Time points for reproducing measured data were chosen from among possible points
when plant parameters were not changing significantly due to the changing amount of water
injection or other reasons. This is because the current evaluation model is based on thestatic heat balance model. Time points after the water injection via the CS started were alsochosen so that the effect of water injection from the CS could be checked.

(1) Unit-2
By referring to temperature changes in Figure 2 (1), time points of August 12, 2011 and
September 12, 2011 were chosen for measured data reproduction before the CS water
injection was started, and September 26, 2001 was chosen for reproduction after the CS
water injection was started.
The temperatures to be reproduced were set as follows. The thermometer at the
feedwater nozzle was the only one available close to the RPV upper surface of interest in
the current model. The temperatures at the feedwater nozzle of Unit-2 changed at a level
lower than the temperatures at the RPV lower part over the whole period of interest forevaluation. For conservative evaluation, temperatures in the RPV upper part were set, by

Attachment 4-10

assuming higher temperatures present somewhere in the RPV upper part. From thetemperature profile of Unit-3 (Figure 2 (2)), the temperature difference between the RPV
shell flange and feedwater nozzle of Unit-3 was taken and added to the Unit-2 feedwater
nozzle temperatures.
As the temperatures of the D/W atmosphere, the temperatures of the D/W HVH return line
were used, which showed stable changes.

(2) Unit-3
Based on the temperature changes in Figure 2 (2), August 12 and August 30, 2011 before
the CS water injection were chosen as the time points for evaluation, and September 12,
2011 after the CS water injection was chosen.
As for the temperatures, temperatures of the RPV shell flange were used as thetemperatures of the RPV upper surface (outer surface of the RPV upper wall) in evaluating
reproducibility of measured data. The temperatures of the RPV shell flange showed thehighest values around the RPV upper part in Figure 2 (2) from August to early September
2011. Higher RPV upper part temperatures predict the higher fraction of uncovered core andlead to a conservative conclusion.
As the temperatures of the D/W atmosphere, the temperatures of the D/W HVH returnline were used, which showed stable changes.

3.2.

Estimation of fractions of the uncovered core and the effective CS contribution

Results of sensitivity analysis are given below, concerning the impact of RPV uppersurface temperatures against the fractions of molten fuel that had been relocated to the PCV
and of uncovered core left at the original position. The water leak fraction (not contributing tocooling) of CS water was set as 20%, 40% or 60%. The results correspond to the fractions
at time points before water injection from the CS started. The fraction of molten fuelrelocated to the PCV was set as 0%, 20%, 40% or 60%.

(1) Unit-2
Figure 3 (1) to Figure 3 (6) give the results of sensitivity analysis of the RPV upper surface
temperatures against the fractions of relocated fuel to the PCV and uncovered core. Fromthese figures, the fractions of uncovered core at the time point chosen to reproduce themeasured RPV upper surface temperatures are obtained as 0.001 to 0.031 on August 12,2011 and 0.008 to 0.027 on September 12, 2011.
By using the parameters obtained above (the fraction of water leaks, the fraction of
relocated fuel to the PCV, the fraction of uncovered core), the fraction of effective cooling by

Attachment 4-11

CS water (the uncovered core direct cooling rate by CS water) was estimated which could
reproduce the RPV upper surface temperatures on September 26, 2011. The results range
from 0.001 to 0.007.
The table below summarizes the results.
Results
Parameters and their value ranges

Water leaks

20 to 60％

Relocation to PCV

0 to 60％

Fraction of uncovered core

Fraction of effective
CS contribution

Aug. 12, 2011

Sept. 12, 2011

Sept. 26, 2011

0.011 to 0.031

0.008 to 0.027

0.001 to 0.007(2) Unit-3
Figure 3 (7) to Figure 3 (12) give the results of sensitivity analysis of the RPV uppersurface temperatures against the fractions of relocated fuel to the PCV and uncovered core.
From these figures, the fractions of uncovered core at the time point chosen to reproduce
the measured RPV upper surface temperatures are obtained as 0.008 to 0.030 on August
12, 2011 and 0.009 to 0.031 on August 30, 2011.
By using the parameters obtained above (the fraction of water leaks, the fraction of
relocated fuel to the PCV, the fraction of uncovered core), the fraction of effective cooling by
CS water (the uncovered core direct cooling rate by CS water) was estimated which could
reproduce the RPV upper surface temperatures on September 12, 2011. The results range
from 0.002 to 0.011.
The table below summarizes the results.
Results
Parameters and their value ranges

Water leaks

20 to 60％

Relocation to PCV

0 to 60％

Fraction of uncovered core

Fraction of effective
CS contribution

Aug. 12, 2011

Aug. 30, 2011

Sept. 12, 2011

0.008 to 0.030

0.009 to 0.031

0.002 to 0.011

Attachment 4-12

3.3.

Estimation of core conditions as of October 10, 2011

The RPV upper surface temperatures and uncovered core fuel surface temperatures as of
October 10, 2011 were estimated, by using the parameters obtained in 3.2.

(1) Unit-2
The results are given in the table below. The RPV upper surface temperatures of 92.2 to
94.9 deg C are a little higher than the measured value of 78.6 deg C. The resulting
uncovered core fuel surface temperatures are 92.7 to 99.3 deg C.
Results
Parameters and their value

RPV upper

Uncovered core

rangessurfacefuel surface

temperaturestemperatures

Water leaks
Relocation to
PCVUpper structure
temperatures

20 to 60％
0 to 60％

Uncovered core(unflooded)

0.008 toEffective CS

0.001 tocontribution

0.007

0.031

92.1 to 94.9
deg C

92.7 to 99.3

92.2 to 96.8

(measured: 78.6

deg C

deg C

deg C)(2) Unit-3
The results are given in the table below. The RPV upper surface temperatures of 71.2 to
81.8 deg C are roughly consistent with the measured value of 71.1 deg C. The resulting
uncovered core fuel surface temperatures are 72.7 to 97.9 deg C.
Results
Parameters and their value

RPV upper

Uncovered core

rangessurfacefuel surface

temperaturestemperatures

Water leaks
Relocation to
PCVUpper structure
temperatures

20 to 60％
0 to 60％

Uncovered core(unflooded)

0.008 toEffective CS

0.002 tocontribution

0.011

0.031

71.2 to 81.8
deg C

72.7 to 97.9

71.4 to 89.0

(measured: 71.1

deg C

deg C

deg C)

Attachment 4-13

4.

Conclusion
Heat balance in the RPV was evaluated using an in-RPV temperature evaluation model.

The results show that the fractions of unflooded fuel (uncovered core) were less than about
3% in both Unit-2 and Unit-3, and the fuel surface temperatures in the unflooded region
were below 100 deg C.
It can be estimated from the current (November 2011) temperature changes at various
points of the RPV and PCV that the inside of the RPV and PCV are being sufficiently cooled
and that uncovered overheated fuel is not present in significant quantities. The evaluation
results of this report can be understood to support the estimation.
End

Attachment 4-14

200
Feedwater nozzle
Temperatures
給水ノズル N-4B 温度
N-4B temperature
RPVまわり温度
around RPV
RPVlower part temperatures
圧力容器下部温度
D/W
HVH
Returnline
temperatures
D/W HVH戻り温度(HVH-16A)
D/W atmosphere(HVH-16A)
D/W雰囲気温度
temperatures
RPVベローシール
RPV Seal bellows

190
180
170

Temperature criterion
判断温度（給水ノズル温度+20℃）(feedwater nozzle temperature+ 20 deg C)

160
150

温度（℃）

140

Evaluation
time point
評価対象日時（8/12）(8/12)

Evaluation
time point (9/12)
評価対象日時（9/12）
Evaluation
time point (9/26)
評価対象日時（9/26）

130

Temperature criterion
判断温度(feedwater
nozzle temperature（給水ノズル温度+5℃）+ 5 deg
C)

120
110
100
90
80
70
60
50
8/3

8/13

8/23

9/2

9/12

9/22

10/2

Date andtime
日時

Figure 2 (1)Temperature changes around the RPV and D/W atmosphere (Unit-2)

140
評価対象日時（8/12）
Evaluation
time point (8/12) 評価対象日時（8/30）
Evaluation time point (8/30)

130Evaluation time point (9/12)
評価対象日時（9/12）

温度（℃）

120

110

100

90

80

給水ﾉｽﾞﾙnozzleN-4B
N4B 温度 temperature
Feedwater
RPV
ｽﾀｯﾄﾞﾎﾞﾙﾄ温度
RPV
studbolt temperatures
Temperatures
RPV
胴ﾌﾗﾝｼﾞ
RPV
shellflange
RPVまわり温度
around RPV
RPVshell
胴ﾌﾗﾝｼﾞ下部温度
RPV
flange lower part
RPVbottom
底部ﾍｯﾄﾞ上部
RPV
heat upper surface
RPV
lower part temperatures
圧力容器下部温度
D/W
Return line temperatures
D/WHVH
HVH戻り温度
D/W atmosphere
D/W雰囲気温度
temperatures
RPV
Seal bellows
RPVﾍﾞﾛｰｼｰﾙ

70
8/3

8/13

8/23

9/2

9/12

9/22

10/2

10/12

Date andtime
日時

Figure 2 (2)Temperature changes around the RPV and D/W atmosphere (Unit-3)

Attachment 4-15

300

RPV上部表面温度（℃）

250

Fraction of uncovered core

0%

0.01

0.04

PCV落下割合 0%
Fraction of uncovered core 20%
PCV落下割合20%
Fraction of uncovered core 40%
PCV落下割合40%
Fraction of uncovered core 60%
PCV落下割合60%
Feedwater nozzle temperatures (+20°C)
給水ノズル温度(+20℃)

200

150

100

50
0.00

0.02

0.03

0.05

0.06

0.07

0.08

0.09

0.10

露出炉心割合（-）
Fraction of
uncovered core (-)

Figure 3 (1)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-2, August 12, 2011, water leak 20%)300

RPV上部表面温度（℃）

250

Fraction of uncovered
core 0%
PCV落下割合
0%
Fraction of uncovered core 20%
PCV落下割合20%
Fraction of uncovered core 40%
PCV落下割合40%
Fraction of uncovered core 60%
PCV落下割合60%
Feedwater nozzle temperatures (+20°C)
給水ノズル温度(+20℃)

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (2)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-2, August 12, 2011, water leak 40%)

Attachment 4-16

300

RPV上部表面温度（℃）

250

Fraction of uncovered
core 0%
PCV落下割合
0%
Fraction of uncovered core 20%
PCV落下割合20%
Fraction of uncovered core 40%
PCV落下割合40%
PCV落下割合60%
Fraction of uncovered core 60%
給水ノズル温度(+20℃)
Feedwater nozzle temperatures (+20°C)

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of
uncovered core (-)
露出炉心割合（-）

Figure 3 (3)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-2, August 12, 2011, water leak 60%)300

RPV上部表面温度（℃）

250

Fraction of uncovered
core 0%
PCV落下割合
0%
Fraction of uncovered core 20%
PCV落下割合20%
Fraction of uncovered core 40%
PCV落下割合40%
Fraction of uncovered core 60%
PCV落下割合60%
給水ノズル温度(+20℃)
Feedwater nozzle temperatures (+20°C)

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (4)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-2, September 12, 2011, water leak 20%)

Attachment 4-17

300

RPV上部表面温度（℃）

250

Fraction of uncovered
core 0%
PCV落下割合
0%
Fraction of uncovered core 20%
PCV落下割合20%
Fraction of uncovered core 40%
PCV落下割合40%
Fraction of uncovered core 60%
PCV落下割合60%
給水ノズル温度(+20℃)
Feedwater nozzle temperatures (+20°C)

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

露出炉心割合（-）
Fraction of
uncovered core (-)

Figure 3 (5)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-2, September 12, 2011, water leak 40%)300

RPV上部表面温度（℃）

250

Fraction of uncovered core

0%

0.01

0.04

PCV落下割合 0%
Fraction of uncovered core 20%
PCV落下割合20%
Fraction of uncovered core 40%
PCV落下割合40%
Fraction of uncovered core 60%
PCV落下割合60%
Feedwater nozzle temperatures (+20°C)
給水ノズル温度(+20℃)

200

150

100

50
0.00

0.02

0.03

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (6)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-2, September 12, 2011, water leak 60%)

Attachment 4-18

300

RPV上部表面温度（℃）

250

Fractionof uncovered core 0%
PCV落下割合0%
Fractionof uncovered core 20%
PCV落下割合20%
Fractionof uncovered core 40%
PCV落下割合40%
Fractionof uncovered core 60%
PCV落下割合60%
RPV胴フランジ温度
RPV
shell flange temperatures

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (7)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-3, August 12, 2011, water leak 20%)300

RPV上部表面温度（℃）

250

Fractionof uncovered core 0%
PCV落下割合0%
Fractionof uncovered core 20%
PCV落下割合20%
Fractionof uncovered core 40%
PCV落下割合40%
Fractionof uncovered core 60%
PCV落下割合60%
RPV胴フランジ温度
RPV
shell flange temperatures

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (8)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-3, August 12, 2011, water leak 40%)

Attachment 4-19

300

RPV上部表面温度（℃）

250

Fractionof uncovered core 0%
PCV落下割合0%
Fractionof uncovered core 20%
PCV落下割合20%
Fractionof uncovered core 40%
PCV落下割合40%
PCV落下割合60%
Fractionof uncovered core 60%
RPV胴フランジ温度
RPV
shell flange temperatures

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (9)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-3, August 12, 2011, water leak 60%)300

RPV上部表面温度（℃）

250

PCV落下割合0%
Fractionof uncovered core 0%
PCV落下割合20%
Fractionof uncovered core 20%
PCV落下割合40%
Fractionof uncovered core 40%
PCV落下割合60%
Fractionof uncovered core 60%
RPV胴フランジ温度
RPV shell flange temperatures

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (10)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-3, August 30, 2011, water leak 20%)

Attachment 4-20

300

RPV上部表面温度（℃）

250

PCV落下割合0%
Fractionof uncovered core 0%
PCV落下割合20%
Fractionof uncovered core 20%
PCV落下割合40%
Fractionof uncovered core 40%
PCV落下割合60%
Fractionof uncovered core 60%
RPV胴フランジ温度
RPV shell flange temperatures

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction
uncovered
Fraction ofof露出炉心割合（-）
uncovered core
core (-)(-)

Figure 3 (11)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-3, August 30, 2011, water leak 40%)300

RPV上部表面温度（℃）

250

Fractionof uncovered core 0%
PCV落下割合0%
Fractionof uncovered core 20%
PCV落下割合20%
PCV落下割合40%
Fractionof uncovered core 40%
PCV落下割合60%
Fractionof uncovered core 60%
RPV胴フランジ温度
RPV
shell flange temperatures

200

150

100

50
0.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0.09

0.10

Fraction of露出炉心割合（-）
uncovered core (-)

Figure 3 (12)Fractions of uncovered core and relocated fuel to the PCV, and RPV uppersurface temperatures (Unit-3, August 30, 2011, water leak 60%)

Attachment 4-21

Attachment-5

Calibration of water level indicators

1.

Principle of water level indicators
A typical fuel range water level indicator used in BWR plants is illustrated in Figure 1. It

measures the reactor water level by measuring pressure difference (Hs – Hr) of two
instrumentation piping systems (reference condensing water chamber side piping, hereafter
described as the reference leg; and reactor side piping, hereafter variable leg) while keeping
the water head Hs in the reference water head at a certain fixed value.
If the water level in the reference leg decreases due to evaporation or other reasons, theHs, which should be constant, is reduced. But, since what is measured (or rather observed)
is only the pressure difference, it cannot be distinguished whether Hs has decreased or Hr
has increased. As a consequence, the apparent reactor water level seems to haveincreased (Figure 2).

2.

Calibration of water level indicators of each unit

2.1.

Unit-1The Unit-1 reactor water level indicators were calibrated on May 11, 2011. The indicatorsthemselves were confirmed to be functioning satisfactorily within the allowable errors. Whenthe instrumentation lines and the reference condensing water chamber were filled with water
thereafter before placing the water level indicators in service, the actual readings werechecked. These readings were off the lower end of the scale.
The differential pressure gauge, temporarily installed after the accident to provide
alternative measurements, was, in addition, off the upper end of the scale. The reactor waterlevel was estimated to be below TAF-500 cm. Figure 3 shows the water level indicator
calibration work.
 Calibration results of water level indicators
Reactorwater level
Input
Outputreference(kPa) Voltage (mV）(cm）

Reactorwater level
converted
(cm)

Error
(％)

0％

－300

－78.5340.7－296.8

+0.4

100％

500

－1.06

199.9

499.5

－0.1

 Readings before and after water filling in the instrumentation lines

Attachment 5-1

Allowable
errors(％)±0.5

Monitoring device

Central monitor(LI-263-122A)

Remote monitor(LT-263-121A)

Temporary differential
pressure gauge

Before filling－170 cm－1.67 m－After filling

Downscale *1

Downscale *1

Overscale *2*1: Below -300 cm (below the lower limit of measurable range)*2: The differential pressure gauge indicated “Off-scale (>100kPa).” This is equivalent in
the water head to about TAF-500 cm or below (the reference value).

2.2.

Unit-2A temporary reading device was set on the normal water level indicator on June 22, 2011,
and the instrumentation lines and the reference condensing water chamber were filled withwater (Figure 4). After water filling, water in the instrumentation lines (reference leg, variableleg) in the PCV behaved as if it had evaporated. This is considered to be due to theatmospheric temperature in the instrumentation lines having been higher than the saturation
temperature. Based on the instantaneous readings of the temporary device after water filling
and the temperature changes thereafter, the reactor water level was estimated to be about
TAF-500cm or below.
The instrumentation lines and the reference condensing water chamber were again filledwith water on October 21, 2011. This was because the PCV temperatures were continuing
to decrease and the temperatures in the PCV were lowered to about 85 deg C. The
pressure difference immediately after water filling was off-scale and therefore, as in June
2011, the reactor water level was estimated to be below the instrumentation lines(approximately TAF-500cm or below). Concerning the variable leg, its pressure readings
indicated a gradual decreasing trend as seen in Figure 5 and therefore the water in thevariable leg is considered to have evaporated. This leads to an estimation that fuel (a heatsource) is present near the variable leg.
Calibration of the normal water level indicator for Unit-2 is not possible yet, since theradiation level at its location is too high and for other reasons.

2.3.

Unit-3

The reactor water level indicator calibration is not possible yet including water filling to
instrumentation lines, since the radiation level at is location is too high and for other reasons.

End

Attachment 5-2

Reference condensing
waterchamber
基準面器

Ｈｓ

Ｈｒ
Differential
pressure
gauge

差圧計

基準面器
側配管

Reference leg

Figure 1

炉側配管
Variableleg

A schematic of reactor water level indicators

Reference condensing
water基準面器
chamber

Reference condensing
waterchamber
基準面器

基準面器
Apparent
water
から水位
level
increasesが減少すwhen
the water
ることで、
level
in the
見かけ上referenceの水位がcondensing
waterchamber上昇
decreased

Actualsituation
炉内水位
of reactor
water
の実際の
level
状況

Water level indicator readings
水位計の指示値
Differential

Differential

pressure
差圧計

pressure
差圧計

gauge

gauge

Actual reactor water level
実際の水位

基準面器
Reference
leg
側配管

炉側配管

Variable leg

Figure 2

Reactor water level indicator readings upon the level decrease in the
instrumentation lines

Attachment 5-3

Figure 3

Calibration work on reactor water level indicators of Unit-1
Reference
condensing
water
基準面器

Reference condensing
water side
基準面器側

reactor炉側
waterlevel side)

原子炉
PCV
格納容器
Pressure indicator

Existing
本設waterlevel水位計
indicator

圧力計(炉側)(reactor waterlevel side)

Pressure indicator (Reference
condensing water side)
圧力計(基準面器側)

Differential pressure
差圧計
gauge

(Temporaryrack)（仮設ラック）Figure 4

A schematic of temporary water level indictors for Unit-2

Attachment 5-4

H-side (variable leg) (MPa)

L-side(reference leg)
(MPa)
Ｌ側（基準面器側）（MPa）

L-sideＬ側

0.230

H-side
Ｈ側

0.220
0.210
0.200

0.090
0.088
0.086
0.084
0.082
0.080
0.078
0.076
0.074
0.072
0.070

0.190
0.180
0.170
0.160
0.150
12:00

Figure 5

12:22

12:45

13:07

13:30

13:52

14:15

14:37

15:00

Changes of pressure indicator readings after water filling at Unit-2

Attachment 5-5

Attachment-6

Radioactivity concentrations in the atmosphere of the primary containment vessel

1.

OutlineGases in the Unit-1 primary containment vessel (PCV) were sampled on July 29, 2011

and again on September 14, 2011 for radioactivity concentration measurements, and thegases in the Unit-2 PCV were sampled on August 9, 2011.
2.

Sampling methods

2.1.

Sampling points of PCV gases
Figure 1 and Figure 2 show schematic drawings of the gas sampling systems. Bothat Unit-1 and Unit-2, a temporary sampling rack was connected to the normal in-PCV
oxygen analyzer rack, gases were sampled at the sampling point in the upper part of
the PCV and returned to the sampling point in the middle.

Isolation valve

In-PCV oxygen analyzer rack (normal)

SP75A

Temporary samplingrack (sampling)

SP75B

Temporary samplingrack (return)

SP75C
X-27
In-PCV oxygen analyzer
sampling point

Figure 1

Unit-1 gas sampling system configuration for one-shot gas sampling

Attachment 6-1

Isolation valve
SP75A

In-PCV oxygen analyzer rack (normal)

X-26

Temporary samplingrack (sampling)Temporary samplingrack (return)

SP75C
SP75B

In-PCV oxygen analyzer
sampling point

Figure 2

2.2.

Unit-2 gas sampling system configuration for one-shot gas sampling

Temporary sampling rack
Figure 3 and Figure 4 show schematic drawings of the temporary gas samplingrack (Type 1 and Type 2). In Type 1 sampling rack, a sampling vessel (about 10 mL)
mounted on the temporary sampling line is isolated with valves on both sides,
removed after gas sampling, connected to a jig to transfer the sampled gas to acontainer while being stirred (diluted with air to 4 times larger volume). The sampled
gas is then injected by a syringe into a gas vial for measurement by the Ge
semiconductor detectors. The system operation procedures specify to purge thesampling line with nitrogen gas before and after sampling.
In Type 2 sampling rack, the sampled gas is led to two water-filled impingers(about 350 mL each) mounted on the temporary sampling line to capture
water-soluble radioactive materials and collect condensate water. A gas vial can beconnected to the sampling line for gas collection. The system operation procedures
specify to purge the sampling line with nitrogen gas before sampling, conduct flow
operations using a bypass line and purge the line with nitrogen gas again after
sampling.

Attachment 6-2

From oxygen
analyzer rack

To oxygenanalyzer rack

Sampling vessel
CoolerFI
Pump
Temporary sampling rack

Figure 3

A schematic of the temporary sampling rack Type 1

From oxygen
analyzer rack

FIQ
FIQ

Pump

To oxygenanalyzer rack

ImpingerGas vial
Temporary sampling rack

Figure 4

2.3.

A schematic of the temporary sampling rack Type 2

Gamma ray nuclide analysis
Sampled gas and condensate water are collected in a gas vial (about 14.1 mL) anda Marinelli beaker (about 500 mL) separately for gamma ray nuclide analysis by Ge
semiconductor detectors. The energy range and resolutions of the semiconductor
detectors are about 50 keV to 2.0 MeV, and about 1.8%, respectively and sufficient

Attachment 6-3

for the present measurement purposes.

3. Results
3.1. Collection of condensate water
Unit-1 in-PCV gas was sampled on July 29, 2011 using Type 1 temporary sampling
rack, Unt-2 in-PCV gas was sampled on August 9, 2011 using Type 2 temporary
sampling rack and Unit-1 in -PCV gas was sampled again on September 14, 2011using Type 2 temporary sampling rack.
In the July 29 sampling (Unit-1, Rack Type 1), the gases flowed at the rate of 0.1
L/min for 2 hours and 1.0 L/min for about 30 min but no condensate water could becollected. This is believed to be because the pump capacity was too small to suction
the condensate water-containing gases to the sampling point before the condensate
water fell onto the lowest part of the piping.
In the August 9 sampling (Unit-2, Rack Type 2), the condensate water was
confirmed to be flowing in the sampling device inlet line (a Teflon tube) whensuctioned at the rate of 10 L/min (vapor components were fully condensed andremained in the piping in plug forms). By adjusting the flow rate, the condensate
water was collected in the impingers and the gases were collected in the gas vial.
In the September 14 sampling (Unit-1, Rack Type 2), the rack used had a larger
pump capacity for higher potential to collect condensate water. At the earliest
sampling on July 29, 2011, the condensate water could not be collected using theType 1 rack. By circulating the gases at the rate of about 10 to 40 L/min for several
tens of minutes, the condensate water was confirmed to be flowing in the sampling
device inlet line (hereafter the Teflon tube), the condensate water was collected in theimpingers and the gases were collected in the gas vial.

3.2. Cs radioactivity concentration
Table 1 gives the measured radioactivity concentrations in the sampled
condensate water, while Table 2 gives the measured radioactivity concentrations inthe sampled gas. The results were used to derive the radioactivity concentrations in
the PCV gases as given in Table 8 by the following conversion equations.
Radioactivity concentrations in the condensate water and gas thus obtained wereconverted to the concentrations in the vapor and gases in the PCV, which are further
converted to the in-PCV concentrations by the weighted average of the vapor fractionin the PCV.

Attachment 6-4

[Conversion equations for in-PCV concentrations]
C1: Cs concentration in vapor

C1 =C water×ρvapor(Tpcv)／ρwater(Tsample)(Eq. 1)C2: Cs concentration in gases

C2 =C sampledgas×Tsample／Tpcv(Eq. 2)

Cpcv: Cs concentration in the PCV

Cpcv =α×C1 +(1－α)×C2(Eq. 3)gaseous phase
where
Cwater: Cs concentration in the condensate water collected (measured) (Table 1)
Csampledgas: Cs concentration in the sampled gas (measured) (Table 2)
Tpcv: Atmospheric temperature in the PCV (Table 3)
Tsample: Atmospheric temperature in the temporary sampling rack (Table 3)ρvapor (Tpcv): Vapor density at temperature Tpcv
ρwater (Tsample): Water density at temperature Tsample (≒1)α: Vapor fraction (Table 6, Table 7)

Extraction

IntakeC sampledgas

Trapping in condensate waterC water
: Cs
C pcv

C pcv

1－α

C2

α

C1

C sampledgas
C water

C1: Cs concentration in vapor
C2: Cs concentration in gases
Cpcv: Cs concentration in the PCV gaseous phase

Attachment 6-5

Cwater: Cs concentration in the condensate water collected
Csampledgas: Cs concentration in the sampled gas
α: Vapor fraction
Figure 5

Conversion to in-PCV concentrations

Attachment 6-6

Table 1

Concentrations in the condensate water collected (measured): Cwater
Radioactivity concentrations (Bq/mL)
Unit-1
Unit-2
Unit-1
Condensate water Condensate water Condensate watercollected (9/14)
collected (8/9)
collected (7/29)
2
2
3.8×10
6.9×10

Nuclide(half-life)
Cs-134(about 2 years)

Cs-137(about 30 years)

Set 1
Set 2

3.8×102

3.1×102

Set 3

3.4×102

4.9×102

Set 1

4.2×102

7.3×102

Set 2

4.4×102

3.2×102

Set 3

4.2×102

5.1×102

―

―

Note

Table 2Concentrations in the gas collected (measured): Csampledgas
Radioactivity concentrations (Bq/mL)
Unit-1
Unit-2
Unit-1
sampled gas
sampled gas
sampled gas
collected (9/14)
collected (8/9)
collected (7/29)
2.8
N.D.*
1.7×101

Nuclide(half-life)
Cs-134(about 2 years)

Cs-137(about 30 years)

Condensate water
not confirmed,
not collected

Set 1
Set 2

3.9

8.2×10-1

Set 3

3.6

8.2×10-1

Set 1

3.4

7.0×10-1

Set 2

5.4

9.6×10-1

Set 3

4.6

N.D.

2.0×101* N.D. means “Not detected,”
Table 3Sampling temperatures
Unit-1
Unit-2
sampled gas
sampled gas
collected (9/14)
collected (8/9)

Temperatures of sampling
environment: Tsample
D/W temperatures*: Tpcv

Unit-1
sampled gas
collected (7/29)

25 deg C

26 deg C

26 deg C

85 deg C

107 deg C

96 deg C* Set as the vapor saturation temperature (at 127 kPaa of D/W pressure) for Unit-2,
by assuming the D/W atmosphere was vapor.
For Unit-1, set as the seal bellows temperature, because the sampling point washigh up in the RPV.

Attachment 6-7

Table 4

Cs concentrations in vapor (converted from concentrations in the condensate
water collected): C1 (from Eq. 1)
Radioactivity concentrations (Bq/mL)
Unit-1 (9/14)

Unit-2 (8/9)

Set 1

1.4×10-1

5.2×10-1

Set 2

1.4×10-1

2.3×10-1

Set 3

1.2×10-1

3.7×10-1

Set 1

1.5×10-1

5.5×10-1

Set 2

1.5×10-1

2.4×10-1

Set 3

1.5×10-1

3.9×10-1

Nuclide (half-life)
Cs-134(about 2 years)

Cs-137(about 30 years)

Table 5

Cs concentrations in gases (converted from concentrations in the gas
collected): C2 (from Eq. 2)
Radioactivity concentrations (Bq/mL)
Unit-1 (9/14)

Unit-2 (8/9)

Unit-1 (7/29)

Set 1

2.4

N.D.

1.4×101

Set 2

3.2

6.4×10-1

Set 3

3.0

6.4×10-1

Set 1

2.8

5.5×10-1

Set 2

4.5

7.6×10-1

Set 3

3.9

N.D.

Nuclide (half-life)
Cs-134(about 2 years)

Cs-137(about 30 years)

Table 6

Set 1

Unit-1 (7/29)

1.6×101

Vapor fractions estimated from total gas volume extracted and condensate water
volume collected (Unit-1, September 2011)
Total gas→→
Condensate
Vapor
volumeTemperature
Converted towater collected
fraction: α
extracted
corrected
vapor volume
490Ｌ
587Ｌ
25mL
7.07E4 mL
0.11

Set 2

396.3Ｌ

475Ｌ

30mL

8.48E4 mL

0.15

Set 3

348.7Ｌ

418Ｌ

95mL

2.69E5 mL

0.39

Table 7

Set 1

Vapor fractions estimated from total gas volume extracted and condensate water
volume collected (Unit-2, August 2011
Total gas→→
Condensate
Vapor
volumeTemperature
Converted towater collected
fraction: α
extracted
corrected
vapor volume
9Ｌ
11Ｌ
550mL
7.32E5 mL
0.98

Set 2

0Ｌ

0Ｌ

160mL

2.13E5 mL

1.00

Set 3

87Ｌ

111Ｌ

150mL

2.00E5 mL

0.64

Attachment 6-8

Table 8

Radioactivity concentrations in the PCV Cpcv (from Eq.3)
Radioactivity concentrations (Bq/mL)

Nuclide(half-life)

Unit-1 (9/14)

Unit-2 (8/9)

Unit-1 (7/29)

average)

1.6

0.44

4.7 to 6.0 (*2)

average)

2.0

0.46

5.5 to 6.9 (*2)

D/W temperatures

85 deg C

107 deg C

96 deg C

D/W pressures

124 kPa[abs]127 kPa[abs]133 kPa[abs]Vapor fraction (*3)

About 47%

About 100%

About 66%

Total
(weighted
(*1)
Total
Cs-137(weighted(about 30 years)(*1)
Cs-134(about 2 years)(*1) Three concentration measurement results of Unit-1 (9/14) and Unit-2 (8/9) were
weight-averaged, that is the concentrations in the condensate water were weighted
according to the amount of condensate water collected, and averaged. The concentrationsin the gas were the simple average of measurements, because the amount collected in thegas vials was kept constant at all sampling times. The sample with radioactivity
concentrations below the detection limits was excluded from the averaging.

(*2) Condensate water was not collected in Unit-1 (7/29). Values here were obtained byreferring to the values of Unit-1 (9/14) and are the values when 40 to 4000 Bq/mL wereassumed for Cs-134 and Cs-137 in the condensate water.

(*3) The vapor fraction was set as the ratio of the saturation pressure at the D/Wtemperature and the D/W pressure itself (Unit-1), and as just-saturated (Unit-2).

3.3. Estimation of the conditions in the RPV from the in-PCV gas analyses
Table 8 shows the radioactivity concentrations (Cs-134 and Cs-137) in the in-PCVgases are higher for Unit-1 than for Unit-2. Simple comparisons are not possible
because of a potential difference in Cs discharge rates due to different vapor
fractions and temperatures in the PCV gases; that said, the results are consistent
with other evaluations suggesting larger damage to the RPV of Unit-1 than the Unit-2
PCV.
The scheme in Figure 6 points to the following. Cs discharged in the process of fuelmelting in the early stage of the accident is considered to have mostly deposited on
the RPV, PCV, in-RPV structures and other structures, or been transferred to theliquid phase. Cs in the PCV gases and in liquid is considered to have been released
from the leaked gases, because significant leaks from PCV gases and liquid have

Attachment 6-9

been suggested by other evaluations. Therefore, the Cs in the PCV gases is
considered to come mainly from additional releases from the reevaporated depositedmaterials.
According to the results of radioactivity analysis of sampled PCV gases, Csconcentrations in the Unit-1 PCV gases are higher than those in Unit-2 PCV gases.
On the other hand, the temperature readings in the PCV are lower at Unit-1 at thetime of sampling than those at Unit-2 at the time of sampling. This may indicate, as
the reason for higher Cs concentrations in the PCV gases of Unit-1 than of Unit-2,
that the amount of Cs discharged from fuel in the early stage of the accident is larger
at Unit-1 than at Unit-2 and therefore the amount deposited, the main source of thecurrent Cs release, is larger at Unit-1 than at Unit-2. The fraction of fuel being
relocated to the PCV is larger for Unit-1 than for Unit-2 which also supports thisestimation.
漏えい
Leak
Reevaporation
再蒸発

Reevaporation
再蒸発

Condensation
Source of
凝縮 FP
放出源
FP discharge
Coagulation
凝集

化学吸着
Chemical
adsorption

Deposit
沈着

熱泳動
Diffusiophoresis

拡散
Diffusion

拡散泳動
Thermophoresis

Resuspension
再浮遊Deposition
重力沈降
by gravity
Gas-liquid
equilibrium
気液平衡

Source of FP discharge

FP放出源

漏えい
Leak

Figure 6

Scheme of the FP discharge mechanism

Attachment 6-10

Attachment-7
Operability checks of local power range monitor (LPRM) detectors (Unit-2, Unit-3)(Attempt to estimate core conditions from LPRM data)

1.

Introduction

Time domain reflectometry (TDR) analyses of signal wave profiles of in-core instrumentation
detectors (local power range monitor (LPRM), startup range neutron monitor (SRNM),
source range monitor (SRM) may locate the position where the instrumentation lines weredamaged (either short-circuited or disconnected). If the damage position is located in thecore region, the damage can be considered to have indicated the fuel damage conditions.
The information may be useful in estimating directly the damaged core conditions. The TDR
analysis of Unit-2 and Unit-3 LPRM signals was conducted in October 2011.
The TDR analyses of Unit-1 detectors were not conducted and are not planned as of now(November 2011) from the following reasons: a superimposed voltage must be applied tothe instrumentation lines to get signals for the TDR analysis but such a voltage cannot beapplied where hydrogen is present at a concentration above its flammable limit; thehydrogen concentration in the Unit-1 reactor pressure vessel (RPV) would be relatively
higher than those in Unit-2 and Unit-3 as of October 2011, because Unit-1 had been cooled
to lower temperatures than Unit 2 and Unit-3 by that time; the TDR signal acquisition at
Unit-1 had been suspended, until the advantages and disadvantages of TDR analysis were
to be evaluated from the results of Unit-2 and Unit-3, but the results of Unit-2 and Unit-3
were found to be unsatisfactory.

2.

Outline of TDR wave profile measurements
The TDR is a method to observe signal echo characteristics, when an instantaneous

signal is sent to the in-core instrumentation detectors from the main control room andechoed back, by the measurement of characteristic impedances. The features ofcharacteristic impedance may help locating the damage points of the instrumentation lines.
Figure 1 illustrates the layout of measurement devices from the main control room to thedetectors and the TDR wave profiles expected to be observed. The instrumentation lines run
from the main control room to the detectors via a containment vessel penetration and an
LPRM connector at the bottom of the reactor vessel (RPV). These components in between
return characteristic echo signals, which help in locating the damaged points of
instrumentation lines.
In the TDR wave profile image diagram, the X-axis corresponds to the distance from themain control room, while the Y-axis corresponds to the impedance. If the line remains intact

Attachment 7－1

as far as to the detectors, a constriction can be identified at the position of the PCV
penetration and the LPRM connector. If the line has a disconnected or short-circuited(lowered insulation) point in between, the impedance is known to change upward or
downward at the point.

Main control room
(MCR)

Reactor buildingCore

Detector connector on MCR side
PCV penetration

TDR analyzer

D/W

TDR Waveform image

PCV penetration

Normal
Tip end of
the detector

Betweendetectorconnector
and detector

Detector connector

Short-circuited

Betweenpenetrationand detector
connector

Figure 1

Disconnected

Disconnected
Short-circuited

Layout of components from the main control room to detectors and TDR wave
profiles expected to be observed (image)3.

Results of TDR wave profile measurements

3.1.

Unit-2

Figure 2 is an image of the TDR wave profiles observed on the Unit-2 LPRM lines.

Attachment 7－2

Unit-2
One line
Disconnected
PCV penetration
About
20m

123 lines

Shortcircuited

PCV penetration

Figure 2

Unit-2 TDR profiles observed on LPRM lines (image)The results indicated that one line had been disconnected and the remaining 123 lines
had been short-circuited (lowered insulation). The disconnection point could be estimated tobe near the PCV penetration by considering the distance from the main control room, while
the short-circuited points could be estimated to be about 20 m beyond the PCV penetrations(near the pedestal, before the LPRM detector at the RPV bottom) (see Figure 3). No direct
hints were obtained concerning the core conditions.

Unit-2
PCV penetration

RPV lower part

Pedestal

D/W

Short-circuited or disconnected

Figure 3

Estimation of short-circuited or disconnected points of Unit-2 from TDR analysis

Attachment 7－3

3.2.

Unit-3Figure 3 is an image of the TDR wave profiles observed on the Unit-3 LPRM lines.
Unit-325 lines

Disconnected

PCV penetration

99 lines
ShortcircuitedsPCVpenetration

Figure 4

Unit-3 TDR profiles observed on LPRM lines (image)The results indicated that 25 lines had been disconnected and the remaining 99 lines had
been short-circuited (lowered insulation). All disconnected or short-circuited points could be
estimated to be near the PCV penetrations by considering the distance from the main
control room (Figure 5). No direct hints were obtained concerning the core conditions.

Unit-3
PCV penetration

RPV lower part

D/W

Pedestal

Figure 5

Estimation of short-circuited or disconnected points of Unit-3 from TDR analysis

Attachment 7－4

4.

Estimation of in-core conditions
The measurements were implemented with the anticipation that the information

concerning the distribution of damaged points in the reactor (disconnected or
short-circuited) of multiple in-core instrumentation lines might give some clues on the
directional dependence of core damage and the integrity conditions of fuel remaining in the
peripheral core region.
However, all instrumentation lines of Unit-2 and Unit-3 indicated their damage points(disconnected or short-circuited) were at points before the lines reached the core region. No
useful information was obtained to evaluate the directional dependence of core damage or
the estimated locations of integral fuel that remained.
The Unit-2 instrumentation cable, which assembled the instrumentation lines, wasestimated to have been damaged near the pedestal, while the Unit-3 cable was estimated tohave been damaged near the PCV penetrations. It is difficult to estimate the fuel conditionsin the reactor. If the damaged fuel had been relocated, it might have damaged the cable in
the RPV lower part. Elaboration is difficult as of now (November 2011) on the quantitativeevaluation of the correlation between the fuel relocation and cable damage or of therelocated position of fuel.

5.

Conclusions
TDR wave profiles of LPRM were measured at Unit-2 and Unit-3 with an objective of

estimating core conditions.
The expectation was to estimate core conditions from the information of locations andother response signals of undamaged detectors. But the anticipated estimation was found to
be quite difficult to achieve, because there were no intact detectors.
After the unsuccessful attempt using the LPRM detectors for this purpose, attempts with
the SRNM or SRM also resulted in similar unsuccessful outcomes. As of now (November2011), conducting additional attempts is considered not to be meaningful due to theincreased radiation exposure for such work.

End

Attachment 7－5

Attachment-8

Condition checks of control rod position indicator probe (PIP) (Unit-1, Unit-3)(Attempt to estimate core conditions from control rod position indicator probe (PIP) data)
1.

Introduction
At boiling water reactors (BWRs) control rods are designed to be partly inserted in various

profiles in the core during operation. Each control rod is equipped with a control rod positionindicator probe (PIP) and its insertion depth is continually monitored during normal reactor
operations. All control rods are known to have been fully inserted in a reactor scram
immediately after the Great East Japan earthquake. In the accident progression process
thereafter, the core was damaged, and part of the core fuel melted and was relocateddownward. Some control rods and control rod drive mechanisms (CRDs) below them might
have been damaged also. If a CRD was damaged, the PIP mounted on it would have beendamaged, too. If a PIP indicates readings other than “fully-inserted” or unrealistic readings, it
might hint at damage conditions at the bottom part of the reactor vessel (RPV). From thisbackground, PIP conditions were checked at Unit-1 and Unit-3 in September 2011.
The condition check of Unit-2 PIPs was not conducted from the following reasons: Unit-2
PIPs needed to be operated from a local control panel, while Unit-1 and Unit-3 PIPs could
be operated in the main control room; the Unit-2 PIP condition check had been suspended,until the advantages and disadvantages of Unit-1 and Unit-3 PIP condition checks were tobe evaluated; but the results of Unit-1 and Unit-3 were found to be unsatisfactory. As of now(November 2011), the Unit-2 PIP condition check is not planned.

2.

Outline of PIP condition checks
A PIP is located near the CRD at the RPV bottom (Figure 1). The control rod position(inserted depth) is detected upon the reaction of the permanent magnet mounted on the
CRD and one of the lead switches vertically arranged on the PIP.
Normally, after the reactor is scrammed, lead switches “00 (Control rod No. 4 positionindication unit at ‘fully inserted’)” and “51 (All control rod in-core position indication unit at‘fully inserted’)” react to the permanent magnet, and other lead switches (for example, “48(Control rod No. 4 position indication unit at ‘fully withdrawn’)” or “20 (Control rod No. 4
position indication unit at ‘partly inserted’)” should not react. If conduction states different
from such normal states can be recognized, clues might be obtained to estimate conditionsin the reactor or the RPV bottom.

Attachment 8-1

In the current check work, conduction states were checked on lead switches “00,” “51,”and “48,” which should have indicated unique positions, and on the lead switch “20” as aposition indicator unit representing other lead switches arranged at mid-positions.

Control rod
drive mechanism
Fully inserted

Permanent
magnetFully withdrawn
Indicator tube
Probe bodyConnector(To rod position indicator system)

Figure 1

Illustrated structures of control rod drive mechanism (CRD) and control rod
position indicator probe (PIP)

3.

Results of PIP condition checks
Conduction states of lead switch contact points of each control rod were checked, which

confirmed conductions at several contact points. The results were categorized into thefollowing.
A: Conductions at 4 contact points
C: Conductions at 2 contact points only at the fully inserted position
D: No conductions at all 4 contact points
In: Conduction only at the inserted position
Dr: Conduction only at the withdrawn position
20: Conduction only at Position 20
00: Conduction only at Position 00?: Conductions at more than 2 contact points other than the fully inserted position

Attachment 8-2

3.1.

Unit-1Figure 2 shows the PIP condition check results of all control rods of Unit-1.

Conditions at the RPV bottom part
Color codes: Conductions at 4 contact points: Conductions at 2 contact points only at
the fully inserted position: No conductions at all 4 contact points

In: Conduction only at the inserted position

Dr: Conduction only at the withdrawn position: Conduction only at Position 20

In

Dr: Conduction only at Position 00: Conductions at more than 2 contact points
other than the fully inserted position

Dr
Dr

Figure 2

PIP conduction states check results (Unit-1)

One PIP marked as “C” on the right side of the figure indicated “normal.” But all other PIPs
were categorized as “unusual.” Several PIP cables penetrated the containment vessel(PCV) in a bundle (PIP cables penetrating the PCV at a particular penetration are colored aseither red, blue, violet or black boxes in the figure). PIPs penetrating the same PCV
penetration seemed to show a similar tendency. For example, all control rods in red boxes in
the upper right were “A,” and all control rods in blue boxes in the upper left were “A,” too.
Control rods in violet boxes in the lower right were randomly categorized, so were thecontrol rods in black boxes in the lower left. From this observation, the information obtained
from the measurement of PIP conditions is more likely to have indicated the conditions at
the position of the PCV penetrations, rather than at the RPV lower part.

Attachment 8-3

3.2.

Unit-3Figure 3 shows the PIP condition check results of all control rods of Unit-3.

Conditions at the RPV bottom part
Color codes
In: Conductions at 4 contact points: Conductions at 2 contact points only at
the fully inserted position: No conductions at all 4 contact points
In

In: Conduction only at the inserted position

Dr: Conduction only at the withdrawn position: Conduction only at Position 20: Conduction only at Position 00: Conductions at more than 2 contact points
other than the fully inserted position

Dr

Figure 3

PIP conduction states check results (Unit-3)

No PIP was marked as “C,” i.e., all control rods were categorized as “unusual.” No
correlation was identified, either, between PIP cables penetrating the PCV at the same
penetration (represented in either red, blue, violet, yellow-green or black boxes in the figure),
as had been observed at Unit-1 (See 3.1).

4.

Estimation of in-core conditions
The overall view of the upper right and upper left of Figure 2 for Unit-1 indicated the highlikelihood of PIP cable damage near the point where the cables were bundled (near the PCV
penetration). This information was not helpful to estimate the conditions at the upper right or
upper left region of the RPV bottom part or the conditions in the reactor. On the other hand,no meaningful features were recognized in the lower half of Figure 2, either, and thedamage conditions at the RPV bottom part could not be estimated. One “normal” PIP could
be confirmed but was not helpful to estimate in-reactor conditions, because no correlation
with the states of PIPs around it was confirmed.
The overall view of Unit-3 reactor in Figure 3 indicated that the right half tended to be
short-circuited, while the left half tended to be disconnected, although these tendencies
were not clearly noticeable. Some PIPs in the right half could be estimated to be

Attachment 8-4

disconnected, and some PIPs in the left half could be estimated to be short-circuited. It isdifficult as of now (November 2011) to estimate conditions in the reactor and RPV bottom
part from the information at this stage. PIPs transfer their signals outside the PCV throughone of five PCV penetrations. A similar tendency was recognized among PIPs penetrating
the PCV at one penetration, but the tendency was not clear enough to draw conclusive
estimations.

5.

Conclusions
PIP conditions of Unit-1 and Unit-3 were checked in order to estimate conditions in thereactor and the RPV bottom.
Attempts were made to estimate the conditions in the reactor and the RPV bottom from
PIP conduction states but they were unsuccessful.
Similar attempts at Unit-2 will result in outcomes similar to those of Unit-1 and Unit-3. It isconsidered as of now (November 2011) that the benefit of additional investigation work at
Unit-2 would not be able to overcome the disadvantage of exposure dose for the work.

End

Attachment 8-5

Attachment-9

Results of condition checks and behavior of drywell (D/W) sump thermometers

1.

Conditions checked(1) Unit-1 (checked on September 23, 2011)
Tag No.

Averageline
resistance (A)Averageline
resistance at thetime of periodic
maintenance (B)A/B

Result of direct
current
resistance (*)

TEtemperatures

TE-2001-412

60.5Ω

76.1Ω

0.795

Insulation
resistancedecrease

36.6℃(2) Unit-2 (checked on November 15, 2011)
Tag No.

Averageline
resistance (A)Averageline
resistance at thetime of periodic
maintenance (B)A/B

Result of direct
current
resistance (*)

TEtemperatures

TE-20-362∞

144.6Ω

－Disconnected－(3) Unit-3 (checked on November 15, 2011)
Tag No.

Averageline
resistance (A)Averageline
resistance at thetime of periodic
maintenance (B)A/B

Result of direct
current
resistance (*)

TEtemperatures

TE-20-362

54.3Ω

60.9Ω

0.892

Insulation
resistance
dropped

39.9℃(*) Criteria for evaluating direct current resistance
Good: 1.1≧R≧0.9, where R=A/B=Resistance measured/Normal resistance
Insulation resistance dropped: R<0.9
Disconnected: R>1.1

Attachment 9－1

2.

Thermometer readings behavior(1) Unit-1

D/W equipment drain sump temperatures

Date and time

Figure 1

D/W equipment drain sump temperature behavior (Unit-1)(Data acquisition started on October 20, 2011 with a digital recorder)(2) Unit-2
Data acquisition failed due to the instrumentation line disconnection.

(3) Unit-3
Reporting is omitted because data have not yet been sufficiently accumulated since thestart of their collection.

End

Attachment 9－2

Attachment-10

Results of condition checks and behavior of PLR pump inlet thermometers

1.

Conditions checked

1.1.

Unit-1 (checked on November 22, 2011)

Tag No.

Averageline
resistance (A)Averageline
resistance at thetime of periodic
maintenance (B)A/B

TE-261-8A

43.5Ω

56.0Ω

0.78

TE-261-8B

43.4Ω

52.7Ω

0.82

1.2.

Result of direct
current
resistance (*)
Insulation
resistancedecrease
Insulation
resistance
droppedTE
temperatures

39.2℃

41.4℃

Unit-2 (checked on November 22, 2011)

Tag No.

Averageline
resistance (A)Averageline
resistance at thetime of periodic
maintenance (B)A/B

Result of direct
current
resistance (*)

TEtemperatures

TE-2-145A

95.5Ω

96.6Ω

0.99

Good

44.7 to 52.8℃

TE-2-145B

93.0Ω

92.4Ω

1.01

Good

50.6℃

Result of direct
current
resistance (*)

TEtemperatures

1.3.

Unit-3 (checked on November 22, 2011)

Tag No.

Averageline
resistance (A)Averageline
resistance at thetime of periodic
maintenance (B)A/B

TE-2-145A

59.2Ω

74.2Ω

0.80

TE-2-145B

58.5Ω

66.4Ω

0.88

Insulation
resistancedecrease
Insulation
resistancedecrease(*) Criteria for evaluating direct current resistance
Good: 1.1≧R≧0.9, where R=A/B=Resistance measured/Normal resistance
Insulation resistance dropped: R<0.9
Disconnected: R>1.1

Attachment 10-1

40.8℃

52.1℃

2.

Behavior of thermometer readings

2.1.

Unit-1

PLR pump (A) inlet water temperatures
PLR pump (B) inlet water temperatures

Date and time

Figure 1
2.2.

PLR inlet water temperature behavior (Unit-1)

Unit-2

PLR pump (A) inlet water temperatures
PLR pump (B) inlet water temperatures

Date and time

Figure 2

PLR inlet water temperature behavior (Unit-2)

Attachment 10-2

2.3.

Unit-3

PLR pump (A) inlet water temperatures
PLR pump (B) inlet water temperatures

Date and time

Figure 3

PLR inlet water temperature behavior (Unit-3)

End

Attachment 10-3

Attachment-11

Contamination of the reactor building closed cooling water (RCW) system

1.

Unit-1 reactor building closed cooling water (RCW) system
Radiation distributions measured on May 9, 2011, at various spots in the Unit-1 reactor

building (R/B) showed high dose rates on the RCW system lines (Figure 1). The RCW is a
closed loop for cooling auxiliary equipment and it is unlikely to be contaminated as high as
several hundreds of mSv/h in normal situations. But the RCW lines were laid widely
throughout the R/B and they cooled the equipment in the containment vessel (PCV), too. As
seen in Figure 2, the RCW line for drain cooling was laid in the equipment drain pit in thelower part of the PCV. Therefore, it was highly possible in Unit-1 that the molten fuel was
relocated to the equipment drain pit and damaged the RCW piping and this caused the high
RCW line contamination. Upon damage of the RCW piping, high dose steam and/or water isconsidered to have transferred to the RCW secondary system piping, accompanying
radioactive materials.
Subsequent measurements of radiation doses were conducted thereafter in the R/B(Figure 3 to Figure 6). High dose rates over 1000 mSv were obtained on Unit-1 R/B Floor 2.
The area of interest was where the RCW heat exchanger had been installed. Extremely highdose rates in the area were considered to have been caused by the large amount ofradioactive materials deposited in the heat exchanger. A heat exchanger is the equipment toexchange heat between the primary side and the secondary side, and therefore the surface
temperatures of RCW primary piping are considered to be lower than those at other parts ofthe primary side piping. Volatile radioactive materials such as Iodine and cesium might havedeposited on the piping surfaces. This might lead to a possibility that the RCW piping
damage at Unit-2 and Unit-3 could be diagnosed by comparing radiation doses in the RCWheat exchanger installation areas.

2.

Unit-2 and Unit-3 RCW heat exchangers
Figure 7 to Figure 13 show radiation dose rate distributions in the Unit-2 R/B and Unit-3

R/B (where measurements were made). The RCW heat exchangers of Unit -2 and Unit-3
were both installed on Floor 2 of each R/B. Figure 10 shows the dose rate distribution on the
Unit-2 R/B Floor 2, while Figure 12 shows that on the Unit-3 R/B Floor 2. The dose rates
around the heat exchangers were both in the range of several tens of mSv/h, not extremely
high dose rates as had been observed at Unit-1. The RCW piping in the Unit-2 PCV and
Unit-3 PCV were likely not to have been damaged.

Attachment 11-1

Unit-1 R/B survey results

May 14, 2011
Tokyo Electric Power Company
Survey date: 16:01 to 17:39 on May 13, 2011 (Dose rates shown in green)
11:32 to 11:58 on May 5, 2011 (Dose rates shown in blue)
72(Two steps down from the top of the stairs)
Unit
04:18 to 04:47 on May 9, 2011 (Dose rates shown in red)79(Top of the stairs)
Floors 1 to 2 platform(Top of the stairs)

Above the platform 5.0
Below the platform 11.5

(Measured with inner door closed)

RCW piping
RCW piping

R/B large equipment
entrance hatch

Figure 1 Unit-1 R/B dose rate survey results

Attachment 11-2

RCW
heat
ＲＣＷ
exchanger
熱交換器

床ドレン
サンプピット機器ドレン
サンプピットFloor drain sump pit

Figure 2

Equipment drain sump pit

Schematic of interface between RCW and equipment drain pit

Attachment 11-3

Unit-1 R/B Floor 4

67
70

22
59
50

20
5

44

21

23

16
19
40
57

Unit: mSv/h
単位：ｍSv/h

142

90
20

4

Unit-1
R/B
Floor ４階
4１号機
Ｒ/Ｂ
Figure 3

Dose rate distribution on Unit-1 R/B Floor 4

Attachment 11-4

3

4

床ガレキ
Rubble100

Unit-1 R/B Floor 3

12
10
17
20
23

30

13

水たまり
Water
puddle200～300

74
149

18

階段下
Bottomof stairs
654
40

11
6.5

159

6.5

25

100

16

10
13

8.5

11

8

5.0
6.0

1.4

1.7

1.2

2.2

2.2

2.7

Unit: mSv/h
単位：ｍSv/h

2.0

4.0

5.0

10
15

233
25

ガレキ
15

１号機
Ｒ/Ｂ
３階
Unit-1 R/B
Floor
3
Figure 4

Dose rate distribution on Unit-1 R/B Floor 3

Attachment 11-5

Stairs
1～3FL
1～3FL
階段

Unit-1 R/B Floor 2

10～15

9
8
24
20

41

46
52

30

100

180

189
240
700

>1000

Unit:
mSv/h
単位：ｍSv/h

１号機
Ｒ/Ｂ
Unit-1
R/B
Floor２階2Figure 5

Dose rate distribution on Unit-1 R/B Floor 2

Attachment 11-6

RCW heatexchangers
installedUnit-1 R/B Floor 1
最大
Max.
79

Unit 1

7.5
4

14

9

Shielding in place since
９／８～遮へい設置
最大 September 8, 2011
Max.

13
4

1～3FL階Stairs段
1～3FL

7

620
1200

10～15

5.0

6.0

最大
Max.
13.6

8
6.5

0.20

0.25

9

:9
Grating:45
グレーチング上：45
Front of staircase ladder :50
2段目はしご前：50
Topof staircase ladder : 60
2段目はしご上：60
Test
tap 14A:70
テストタップ14A：70 :90 to 100
MO-001A

7

Enlarged
view
詳細図

10

20

20
above
above
上部上部
2.5 50 100
above
上部
9
7
25
1.5
above
10
上部 above上部
35
40 130

2 2 2
2
above
上部
上部上部
2.5 above上部
4.5 4 4.5上部
above 43 3.5
13
上部
7 above
17

10
5

10
38

0.36

0.75

0.90

0.45

1.1

0.65

0.33

0.23

1.3
4.75

2.85

13

1.5

45
Immediately
ﾌｪﾝﾈﾙ直上 Floor penetration
above funnel 床貫通部
2200
2200
47004700

90
8.5
1.6

12

12

121

1.6
90

0.1

220

54

247

300
290

22

220
800

48

127

１号機
Unit-1
R/BＲ/Ｂ
Floor１階
1
Figure 6

Dose rate distribution on Unit-1 R/B Floor 1

Attachment 11-7

1100 1280
Max.
最大
3000

1820
138

26

単位：ｍSv/h
Unit:
mSv/h

30

4

上
部

6

25

"above" means
上部は高さ2.2m above from
２．２ｍの値the floor.

MO-001A：90～100

6

0.7

1700

5 Staircase
ladder
1段目はしご上：9

10

7.5

above
上部

7.5～8.0600

260Unit-2 R/B Floor 5

114
190
207

250

単位：ｍSv/h
Unit: mSv/h

２号機
Ｒ/Ｂ
５階
Unit-2 R/B
Floor
5
Figure 7

Dose rate distribution on Unit-2 R/B Floor 5

Attachment 11-8

Unit-2 R/B Floor 4

45

Unit: mSv/h
単位：ｍSv/h

Unit-2
R/B
Floor４階
4２号機
Ｒ/Ｂ

Figure 8

Dose rate distribution on Unit-2 R/B Floor 4

Attachment 11-9

Unit-2 R/B Floor 3
25

11

11
13

38

10
13
12
29

18
40
24

単位：ｍSv/h
Unit:
mSv/h

17

２号機R/B
Ｒ/Ｂ
３階
Unit-2Floor
3
Figure 9

Dose rate distribution on Unit-2 R/B Floor 3

Attachment 11-10

RCW heatexchangers
installed

Unit-2 R/B Floor 2
Platform
踊り場
13

55.6
17

18

16

Max.
最大
50

31

33

40
43
97.2

30.4
38.1

67.6

79.1

Unit:
mSv/h
単位：ｍSv/h

Unit-2
Floor２階
2
２号機R/B
Ｒ/Ｂ
Figure 10

Dose rate distribution on Unit-2 R/B Floor 2

Attachment 11-11

27
52
20

31.1

Unit-2 R/B Floor 1

Platform under
階段下踊り場the staircase

30
Above the TIP cell

25
Platform
踊り場 3030
On the 壁面
wall 15
15

25

20

6.7

20

17

27

25

20

25
25
3.0

9.0

21

9

35
17
25

20
14

15
コンジットパイプ
X-53

Front of sump rack

サンプラック前28

50

Large component entrance hatch Floor 1

Front of 計装ラック前
instrumentation rack

45

35

20

20

22
X-6ペネ上部

Front アクセス用
of access ladder30
はしご前：3036.0
25Large component entrance hatch Floor 2

10

Front of計装ラック前
instrumentation rack

19

Unit:
mSv/h
単位：ｍSv/h

50
60

7

20～25

Unit-2Floor
1
２号機R/B
Ｒ/Ｂ
１階
Figure 11 Dose rate distribution on Unit-2 R/B Floor 1

Attachment 11-12

RCW heatexchangers
installed

Unit-3 R/B Floor 2

98
98

90
54
52
75

22

32

50

45

弁近傍
45
Near
the valve
70

45

50operation 50
Top操作用階段上端
of the staircase for
On床面
the floor
60

60

48

170
Near弁ハンドル近傍
the valve handles
170
Shielding box surface 280
遮へいボックス 280

Unit: mSv/h
単位：ｍSv/h

３号機
２階 2
Unit-3Ｒ/Ｂ
R/B Floor
Figure 12

Dose rate distribution on Unit-3 R/B Floor 2

Attachment 11-13

52

Unit-3 R/B Floor 1

27
34
20

27

430

53

37～50

38

45～74
64～66

50

38

190

46～15595～135

Near the弁近傍 37
valve

620
41

43

43～45

150～224
215

78

96

142

41.5

29.3
30

60
63

24.8

96
42

38

Top half/Bottom half of
instrumentation
rack
計装ラック
上段～下段

100

80～90

42

32
Max.
最大
122

24

30～40

単位：ｍSv/h
Unit:
mSv/h
Max.
最大
388

87

96

20

27

51

３号機
Unit-3
R/BＲ/Ｂ
Floor１階
1
Figure 13

Dose rate distribution on Unit-3 R/B Floor 1
End

Attachment 11-14

Attachment-12

Impacts of core-concrete reactions on the reactor containment vessel
When the damaged core melts core internal structures, melt stacks on the RPV bottom
as fuel debris, and it is relocated to the reactor containment vessel (PCV) pedestal, thepedestal can be damaged due to the so-called molten core-concrete reactions (MCCI), in
which the fuel debris reacts with the pedestal concrete.
This document presents review results of the impacts of MCCI on the PCV at each of
Unit-1 to Unit-3.

1.

Molten core-concrete reactions (MCCI)(1) Outline of a core-concrete reaction
A core-concrete reaction is a chemical reaction in which concrete is decomposed whenheated up above its melting point upon contact with high temperature fuel debris. Whendecomposed, the concrete generates hydrogen, carbon dioxide and other gases, while thefuel debris dissolves the residual concrete and erodes the concrete. When the fuel debris is
not sufficiently cooled, and the decay heat exceeds the amount of heat to be released fromits surface, the excess heat is absorbed by the concrete. When the concrete temperatures
exceed its melting point, the erosion starts.
The reactions decline with time and eventually cease in a limited time in a limited erosion
volume as the decay heat decreases with time and the core-concrete contact area increases
in inverse proportion to the progress of erosion.

(2) Phases of core-concrete reactions development

Upon relocation of molten fuel to the reactor containment vessel (PCV), it will spreadon the pedestal floor if the fluidity is maintained, part of the molten fuel further will leakout through pedestal slits and solidify as flat lumps with a large surface (Figure 1).
If there are openings on the pedestal floor, such as an equipment drain sump pit, thefuel debris may clog them heavily (Figure 2).
If water is retained on the bottom of the PCV, the molten fuel will solidify, being cooled
upon contact with water, into many small lumps. Thus, there are large uncertainties in
configurations and distributions of fuel debris, once the molten fuel is relocated to the
PCV.
Large uncertainties also exist in the heat transfer from the fuel debris to water. Shortly
after relocation, the fuel debris is considered to contact water with a solidified crust in

Attachment 12-1

between (Figure 3). Carbon dioxide and other gases would be generated during theconcrete erosions, and these gases would build up and would crush the crust. The
molten fuel underneath can erupt above the crusts and form fine grains in the water pool.
Part of the coolant water may flow downward through the crushed crust and cool thefuel debris (Figure 4).
Various assumptions are thus needed, and therefore large uncertainties would be
inevitable, in estimating the erosion conditions in the PCV pedestal.

Pedestal
ペデスタル

Pedestal
ペデスタル

About
約2.６m

About
約7.6m

About
約2.６m

ピット
Pits
(1.2m
deep)
(深さ1.2m)

Figure 1 Estimated configuration of fueldebris relocated to the PCV pedestal(fluid fuel debris spreads widely)

-

About
約7.6m

ピット
Pits
(1.2m deep)(深さ1.2m)

Figure 2 Estimated configuration of fueldebris relocated to the PCV pedestal(fuel debris clogs the pits)

Debris configuration
 In a spread fuel debris geometry, the surface area per unit decay heat
enlarges and the core-concrete reactions cease earlier.
 In a stacked fuel debris geometry at one place, the surface area perunit decay heat decreases and the core-concrete reactions ceaseslowly.

Attachment 12-2

Eruption of

Boiling andクラスト上面
radiationatop
the crust
沸騰及び放射溶融デブリ
molten fueldebris
の噴出

Water

Wa

concrete

Water

クラストへのpenetration
into
crust
浸水

crust

Crust
クラスト
formationの形成

MoltencoriumUnsteady heat

コンクリート内
conduction in
concrete
非定常熱伝達Heat convection inConcrete erosion
andコンクリートのgas generation

デブリ/コンクリート
debris-concrete
mixture
対流熱伝達

浸食及び気体発生

Figure 3

Geometrical configuration of fuel debris – concrete – coolant water (image)

Coolant water
Crust
DebrisConcrete(a) Water infiltration into crusts

Figure 4-(b) Debris eruption(c) Crust crushing

Heat transfer from fuel debris to cooling water (image)Heat transfer
 Coolant water from above may solidify the top part of the fuel debris into
crust, which may inhibit heat transfer and limit heat removal.

 Internal pressures due to gases generated in core-concrete reactions may
enhance heat transfer in the debris and at the boundaries.

Attachment 12-3

2.

Evaluation of MCCI impacts(1) Conditions for evaluating concrete erosion depths
Several assumptions are inevitable in evaluating concrete erosion depths because of
unknown debris configurations and cooling environment uncertainties as described in theprevious section. The analysis results can vary significantly depending upon theseassumptions.
The following are the model and conditions set for the analysis.
①

Outline of analysis model
A MAAP built-in sub-program “DECOMP” for core-concrete reaction analyses wasused in evaluating concrete erosion depths. The analysis model is outlined below.
-

Debris compositions (those of molten fuel and in-core structures dissolved by thefuel debris) were taken from the MAAP results.

-

The ORIGEN2* model was used for the fuel debris decay heat. The MCCI was set
to have started at the time when the reactor vessel (RPV) had been damaged in
the MAAP analysis and the decay heat attenuation thereafter was considered.
* Revised version in 1980 of the original nuclear fuel burn up calculation model
ORIGEN that was developed in 1970 by the US Oak Ridge National Laboratory.

-

Heat generated when the zirconium was oxidized, which the fuel debris had
dissolved before it was relocated to the pedestal, was taken into account.

-

The fuel debris was assumed to spread uniformly on the pedestal floor, but part of
the fuel debris was assumed to flow out to the drywell floor through slits. The fueldebris also flowed into the equipment drain sump pits as well as the floor drainsump pits and piled up. Figure 5 illustrates the analytical model of the configurationof fuel debris that had been piled up in the drain sump pits.

-

The fuel debris piled up in the drain sump pits was assumed to be always covered
with the coolant water, and the heat removal (heat flux) by the water was assumed
to be constant at 125 kW/m2, by referring to the MCCI test data (debris coolingtests under the condition of atmospheric pressure and silicic acid base concrete),
which had been obtained by the OECD.

-

The fuel debris layer in the drain sump pitsThe fuel debris was assumed to form a homogenous pool.

Crust layers were assumed to be formed on the top and bottom (side) of thefuel debris pool.

-

The crust layers in the drain sump pits

Attachment 12-4

The changing rates of crust layer thicknesses were estimated from the energybalance of the crusts (heat transfers from the fuel debris pool and to thecoolant water or concrete).

-

The concrete erosion in the drain sump pitsTemperature distributions from the concrete surface into its depth direction
were calculated in a one-dimensional heat convection model.

Erosion was assumed to start as soon as the concrete temperatures exceed
its melting point at 1500 deg K.

The erosion depths were estimated from the heat balance between the
inflows from the fuel debris and losses by the latent heats of concrete
decomposition and dissolving.

The interface area of fuel debris and concrete was assumed to increase(Figure 6).

-

Heat conduction model

Between the top crusts and coolant water: heat removal from the fuel debris(constant heat flux), constant heat transfer area (the drain sump pit
cross-section)In the crust: a parabolic temperature distribution assumed



Between the fuel debris pool and crust: heat transfer by convection

Between the fuel debris pool and concrete: the total of heat convection fromthe fuel debris pool to the crust layer and the decay heat within the crust

Coolant water

Top crust

Molten fuel debris
Off-gas bubbles
Bottom crust
Concrete
Figure 5

A schematic of fuel debris configuration in the analysis model

Attachment 12-5

Coolant water↓Erosion boundary
Concrete
Figure 6

A schematic of concrete erosion in the analysis model②

Condition settings for analysis-

Fraction of core fuel having been relocated to the PCV pedestal
The amount of relocated fuel debris is not certain as of now (November 2011). In
the analysis, the following values were assumed based on the maximum fractions
that have been obtained in the MAAP analysis. Therefore, the evaluation should
lead to conservative results.

-Unit-1: 100%

Unit-2: 57%



Unit-3: 63%

Decay heat
An assumption was made that the volatile fission products (volatile FPs) had beendischarged from the fuel debris before the fuel debris had been relocated from the
RPV to PCV pedestal and that, therefore, the decay heat of volatile FPs had notcontributed to MCCI (i.e., 20% of the decay heat was assumed to have attenuatedbefore the fuel debris relocation).

-

Initial water inventory in the pedestal before fuel relocation
Seal water was supplied to the mechanical seal component of reactor coolant
recirculation pumps to protect reactor water from flowing out. This seal water was
supplied by the control rod drive mechanism system (CRD system). Upon the station
blackout, the CRD system stopped its functions and part of the reactor water wasconsidered to have flowed out via the mechanical seal components.
The reactor water, which flowed out from the mechanical seal component, traveled

Attachment 12-6

to the equipment drain sump pit in the PCV pedestal via the drain line. Excessive
water over the equipment drain sump pit capacity was accumulated in the pedestal,part of which might have further traveled to the drywell floor though slits until the waterlevel reached the lower end of vent tube to the suppression chamber.
If a sufficient amount of water was accumulated by the time of fuel debris relocation
to the pedestal, part of the fuel debris would be formed into fine grains when contactedwith the water, by which the fuel debris cooling would advance. The following
assumptions were made to evaluate the advanced cooling by fine grain formation.
(There are other cooling processes for the fuel debris, but only the cooling effect byfine grain formation of fuel debris was considered in the current evaluation forconservative results.)
Unit-1: the amount of water was assumed to be insufficiently accumulated in thepedestal to form fine grains of fuel debris, because the MAAP analysis had
predicted a relatively short time for the fuel debris to be relocated to the pedestal.
The situation would change little even if the amount of water that had been
accumulated in the equipment drain sump pit during the normal operations was
taken into account.



Unit-2: the amount of water was assumed to be sufficiently accumulated to reachthe level at the bottom end of the vent tube to the suppression chamber, because
the MAAP analysis had predicted a relatively long time for the fuel debris to be
relocated to the pedestal. Two paths of fuel debris relocation were assumed toevaluate the degree of fine grain formation of fuel debris. One was a path for thefuel debris to be relocated through the instrumentation line penetrations, and theother path was through the CRD penetrations. The degree of fine grain formation
was evaluated by using the Ricou-Spalding correlation equation and otherrelevant equations*. The fuel debris not having been formed into fine grains wasused to evaluate concrete erosions.
*The fuel debris was assumed to be relocated in a cylindrical jet form in the
diameter of either the instrumentation penetration or CRD penetration. Part of thejet would disperse into fine grains upon rushing into the pedestal water surface.
The equivalent jet diameter would decrease thereafter. The fraction of fine grain
formation was estimated from the different jet diameters at the time of entry into
the pedestal water and at the time of arriving at the pedestal floor. Specifically, thefractions of fine grain formation per unit surface area were obtained by the
Ricou-Spalding

correlation

equation

(F.B.

Ricou

and

D.B.

Spalding,"Measurements of Entrainment by Axisymmetrical Turbulent Jets," Journal of

Attachment 12-7

Fluid Mechanics, Vol.11, 1961). The jet diameter on arriving at the pedestal floor
was estimated by integrating the results of fractions of fine grain formation along
the relocation path of the fuel debris to the floor. Figure 7 illustrates an image offine grain formation in the water.
-

Unit-3: same as in Unit-2.

Fuel debris sediment
Certain uncertainties are considered to be inevitable in estimating the situation of
relocation development of fuel debris from the RPV bottom to the PCV and theconfiguration of fuel debris sediment.
In the current evaluation, the configuration of sediment on the pedestal floor andthe drywell floor was assumed as follows on a prerequisite that the fuel debris would
have been relocated from the RPV bottom (Figure 8).
The fuel debris was assumed to maintain its fluidity due to its high temperatures
while being relocated down in the pedestal water, but to lose its fluidity gradually byheat radiation while spreading from the pedestal floor to the drywell floor. The fueldebris fluidity would have been further lost by the injected water for cooling.
The fuel debris would spread isotropically upon relocation on the pedestal floor.

Part of the fuel debris that had been relocated to the pedestal floor would further
flow out to the drywell floor through slits. The fuel debris could have damaged thePCV steel plates (the so-called PCV shell attack) depending on the fuel debris
spillage conditions to the drywell. But no plant parameters of Unit-1 to Unit-3
indicated until now (November 2011) have hinted at changes due to the PCV shellattack.

The fuel debris that had spilled to the drywell was assumed to spread radially withan angle of 130 degrees. This assumption was based on research results.



The fuel debris would melt the metallic covers of the equipment drain sump pit andfloor drain sump pit in the pedestal and flow into these pits. For conservativeevaluation, not only the fuel debris sedimented on the pit covers, but the part offuel debris sedimented on the pedestal floor within the fan-shaped 90-degree areaaround the pit was assumed to have flowed into the pits, too.

The sediment thickness of fuel debris in the drain sump pits should be larger thanthat spread uniformly on the pedestal floor or drywell floor. The concrete erosion
depths were evaluated, therefore, on the concrete around the fuel debris sedimentin the drain sump pits.

Attachment 12-8

Jet diameter (when started)

Fuel debris
jetPedestal water
Water depth
Dispersed fine grains

Figure 7

Jet diameter (after arriving at the pedestal floor)
A schematic of fine grain formation

Attachment 12-9

Pedestal (P/D) floor

Drywell (D/W) floor
Drain sump pits

130°

Fuel
debris flowed into pits
サンプへ流入するデブリ
Fueldebris spreading on
P/D，D/W床に拡がるデブリ
the P/D and D/W floors

Part of the fuel debris formed
fine grains (Unit-2, Unit-3)
Figure 8

Fuel debris sediment configuration (image)(2) Results of erosion depth evaluation-

Unit-1
The results of erosion depth estimation are shown in Table 1, Figure 9 and Figure 10 and

Figure 11.
Fraction of fuel relocation

100 %Fuel debris sediment thickness

0.81 mErosion depths

0.65 mTable 1

Unit-1 concrete erosion depth evaluation results

Attachment 12-10

Time after the accident (d)Figure 9

Progression of concrete erosion with time (Unit-1)

1.2 m
1.45 m0.81 m0.65 m0.65 mFuel debris

1.02 mErosion depths

0.65 m
PCV steel plate
Note: Cross-sectional geometry
assumed to be round-shaped

1.02 mFigure 10

A schematic of concrete erosion (vertical view) (Unit-1, Fuel relocation 100%)

Attachment 12-11

Pedestal outer surface

Inner skirt
0.65m

Pedestal inner surface

1.45m
2.5m
3.1m
0.75m

3.7m

Cross-sectional configuration at
the erosion depth
Drain sump pit

Slit

0.65m

1.45m
2.5m
3.1m

Cross-sectional configuration at
the erosion depth

0.75m

3.7m

The erosion volume surface is considered to become round-shaped as the sump pit walls
became eroded by the fuel debris sediment in the drain sump pits.
The actual cross-sectional configuration of the erosion volume will be a geometry between
the top and bottom figures.
Figure 11

Schematic of concrete erosion (plane view) (Unit-1, Fuel relocation 100%)
Attachment 12-12

-

Unit-2
Table 2, Figure 12 and Figure 13 show the results of erosion depth estimation.
Fraction of fuel relocation

57%
Instrumentation line
penetration
0.62

CRD penetration

Debris sediment thickness

0.20 m0.40 mErosion depth

0.07 m0.12 mRelocation path
Fraction of fine grain formation

Table 257%

0.27

Unit-2 concrete erosion depth evaluation results
57% relocation throughinstrumentation line
(RPV damaged in MAAP analysis)
Relocation through CRD line

Time after the accident (d)Figure 12

Progression of concrete erosion with time (Unit-2)

1.2 m
1.45 m0.2 m/0.4 mFuel debris

0.07 m/0.12 m0.07 m/0.12 m0.07 m/0.12 mFigure 13

Note: Cross-sectional geometry
assumed to be round-shaped

A schematic of concrete erosion (Unit-2, Fuel relocation 57%)

Attachment 12-13

-

Unit-3
Table 3, Figure 14 and Figure 15 show the results of erosion depth estimation.
Fraction of fuel relocation

63 %Instrumentation line
penetration
0.56

CRD penetration

Debris sediment thickness

0.31 m0.53 mErosion depth

0.13 m0.20 mRelocation path
Fraction of fine grain formation

Table 3

63 %

0.25

Unit-3 concrete erosion depth evaluation results
63% relocation throughinstrumentation line
(RPV damaged in MAAP analysis)
Relocation through CRD line

Time after the accident (d)Figure 14

Progression of concrete erosion with time (Unit-3)

1.2 m
1.45 m0.31 m/0.53 mFuel debris
0.13 m/0.20 m0.13 m/0.20 m0.13 m/0.20 mFigure 15

Note: Cross-sectional geometry
assumed to be round-shaped

A schematic of concrete erosion (Unit-3, Fuel relocation 63%)

Attachment 12-14

3.

Conclusion
The erosion depths of pedestal concrete due to core-concrete reactions were evaluatedfor Unit-1 to Unit-3. The deepest erosion depths under the assumptions set in the current
study were 0.65 m at Unit-1, 0.12 m at Unit-2 and 0.20 m at Unit-3.
Current results indicate that the largest erosion occurred at Unit-1 but its erosion depth
was still less than the depth to reach the PCV steel plate (1.02 m), i.e., the fuel debris
eroded the pedestal floor but could remain inside the PCV.
There were other possible development processes in fuel debris behavior when it isrelocated, which were not dealt with in the current quantitative evaluation. For example, amodel (*) was not assumed, in which the crust was crushed by the gases generated in thefuel debris while being cooled and the molten fuel debris erupted above the crust. In thiscase, the cooling effects would be significant. The consequences of concrete erosion would
be milder than the results in the current evaluation, i.e., the current evaluation results would
be conservative.
Other forms of development which may affect the fuel debris cooling, or the debris
sediment depths include: fuel debris cooling by the water having been present in theequipment and floor drain sump pits at the beginning of accident; debris sediment thickness
decrease by possible debris leaks to drain pump pits via a connection pipe from theequipment and floor drain sump pits; heat removal by the melting of CRD mechanisms andother components in the pedestal or fuel debris heat density decrease, etc.
(*)

The fuel debris when erupted solidifies in crushed stone forms. Because of their unevensurfaces, the solidified fuel debris is well cooled. In parallel, coolant water infiltrates the crustthrough the crushed gaps and contact fuel debris, advancing cooling.

The Japan Nuclear Energy Safety Organization (JNES) earlier also evaluated the impacts
of MCCI on the PCV. These results concluded, despite different assumptions set in theanalysis such as the debris configuration when sedimentation occurred, that the fuel debris
had remained in the PCV, although it eroded the pedestal floor. (See the Annex to thisdocument for major differences in analysis conditions.)

End

Attachment 12-15

Annex 1

Comparison of the current IAE evaluation results of impacts of core-concrete reactions on
the reactor containment vessel with the JNES evaluation results

JNES issued the following evaluation reports concerning the erosions of PCV (pedestal)by the fuel debris.
-

Report (1): Issue No.6 to be investigated, Core-concrete interactions(CCI), March 25, 2011-

Report (2): Possibility of core-concrete reactions (MCCI) and theirimpacts, April 6, 2011-

Report (3): Possibility of core-concrete reactions (MCCI) and theirimpacts (Part 2), April 7, 2011-

Report (4): Possibility of core-concrete reactions (MCCI) and theirimpacts (Part 3), April 13, 2011

Key conditions for analysis are compared below between the current IAE analysis and the
JNES analysis by referring to these published reports.
○

Fraction of fuel relocation
IAE set the fraction of core damage, based on the MAAP analysis, as the maximum
amount of fuel debris (100% core at Unit-1, 57% core at Unit-2 and 63% core at Unit-3)
assumed in the case of the RPV being damaged.
JNES set 100% core for all units, Unit-1 to Unit-3, as the amount of fuel debris except inReport (2). (In Report (2), 70% core for Unit-1, and 30% core for Unit-2 and Unit-3 wereassumed).) JNES estimated larger erosion depths for Unit-2 and Unit-3.

○

Conditions of fuel debris sediment on the pedestal floor
IAE set the fuel debris sediment thickness by assuming that the fuel debris having been
relocated from the RPV bottom would uniformly spread on the pedestal floor andthereafter partly further spread out to the drywell floor through pedestal slits or partly flow
into the equipment and floor drain sump pits on the pedestal floor.
JNES set the fuel debris sediment thickness by assuming that the fuel debris having
been relocated from the RPV bottom would uniformly spread on the pedestal floor (drainsump pits were not considered).
 JNES assumed thicker debris sediments on the pedestal floor, while IAE assumed
sediments in the drain sump pits. Both assumptions are conservative in evaluating

Attachment 12-16

erosion depths.
○

Configuration of fuel debris
IAE assumed that, at Unit-2 and Unit-3, part of the molten fuel debris would be formed
into fine grains, before it reached the pedestal floor, in the water layer in the pedestal,
which had been accumulated with the water having leaked out from the reactor throughthe recirculation pump mechanical seal component. The MCCI was evaluated by using
the fuel debris which had not been formed into fine grains. As for Unit-1, the amount ofwater was assumed to be insufficiently accumulated in the pedestal to form fine grains offuel debris, because the MAAP analysis had predicted a relatively short time for the fueldebris to be relocated to the pedestal.
In the JNES evaluation, the molten fuel debris was assumed to form a pool on thepedestal floor, although JNES acknowledged the cooling effect of crushed debris in the
coolant water.
 The JNES assumption was more conservative for evaluating the erosion depths at
Unit-2 and Unit-3.

○

Decay heat
Both IAE and JNES assumed the attenuation of the volatile FP contribution to the
decay heat as 20% (the volatile FP decay heat was subtracted from the fuel debris decay
heat as the heat source for the MCCI, assuming that the volatile FPs had volatilized prior
to the MCCI).
IAE used ORIGEN2 in evaluating decay heat. The MCCI was set to have started at the
beginning of the accident (when the RPV was damaged). The heat generated in thezirconium oxidation was also added, in addition to the decay heat, to the heat source for
the MCCI.
JNES used the ANSI/ANS5.1-1979 or May Witt data for evaluation. The MCCI was
assumed to start 20 days after the accident. The heat generated in zirconium oxidation
was not taken into account (i.e., the zirconium was assumed to have been oxidized
completely before the core-concrete reaction started).


IAE assumed core-concrete reactions to have started earlier after the accident. IAE

added the heat generated by the zirconium oxidation to the heat source for core-concrete
reactions. For these reasons, IAE assumptions are more conservative for erosion depths
evaluation than the JNES assumptions.
○

Cooling model
Both IAE and JNES assumed that the molten fuel debris contacted coolant water at its
top and heat was removed via the crust having been formed on the boundary andcontacted concrete at its bottom and heat was discharged for concrete erosion via the

Attachment 12-17

crust having been formed on the boundary.
JNES assumed a model, trying a realistic modelling, in which the upper crust would be
crushed by gas pressures having been generated in the course of MCCI and the fueldebris would erupt above the coolant water.


IAE did not assume the fuel debris would erupt. This should have resulted in moreconservative concrete erosion.
○

Erosion depths
JNES Reports (1) to (4) do not specify the difference in concrete erosion depths
depending upon different analysis conditions, but Report (4) specifically concluded theerosion depth as “about 1 m” and the fuel debris would remain in the PCV. The IAE resultin the current study predicted the erosion depth at Unit-1 was about 0.65 m and that thefuel debris would remain in the PCV.

Comparisons of analysis conditions are summarized below.

Attachment 12-18

Fractionofrelocation

fuelDebris sediment
Debris configuration
Decay heat (source)
Volatile FP decay
heat attenuation
Timing of MCCI
initiation
Heat of Zr oxidation
Erosion depths(Cooling effect oferupted fuel debris
not considered)
Erosion depths

Table 4 Comparison of analysis conditions in the IAE and JNES evaluations
JNES Report (1)JNES Report (2)JNES Report (3)JNES Report (4)IAE
March 25, 2011
April 6, 2011
April 7, 2011
April 13, 2011100%
1F1: 70%
100%
100%
1F1: 100%
1F2,3: 30%
1F2: 57%
1F3: 63%
P/D floor only
P/D and D/W floor
P/D floor only
P/D floor only
P/D floor onlyFine
grainsnot Finegrainsnot Finegrainsnot Finegrainsnot Fine grains formed informedformedformedformed
1F2 and 1F3
Not specified
ANSI/ANS5.1 -1979
May Witt
Not specified
ORIGEN2
Considered

Not considered

Considered

Considered

Considered

Immediately
after
RVP damage
Not considered
1F1 to 1F3:
Fuel debris is cooledandsolidifies
beforeerosionstarts.

20 days after theaccident
Not specified
1F1: 0.92m
1F2: 0.07m
1F3: 0.07m

27 days after theaccident
Not specified
1F1: 1.8 m
1F2: Not ceased byday 10
1F3: Same as 1F2

27 days after theaccident
Not considered
1F1 to 1F3: about
1m (No specificvalues reported forrespective units)

Immediately
RVP damage
Considered
1F1: 0.65m
1F2: 0.07m*1
0.12m*2
1F3: 0.13m*1
0.20m*2

1F1: 0.48m*3
0.63m*4(Cooling effect of－－
1F2: 0.7 m*3
erupted fuel debris
1.1 m*4
considered)
1F3: Same as 1F2
Note: 1F1, 1F2 and 1F3 are named as Unit-1, Unit-2 and Unit-3, respectively, in other places in the report.

Attachment 12-19－

Remarks

after*1: Fuel relocated via
instrumentation line
*2: Fuel relocated via
CRD penetration*3: Debris eruption
coefficient E=0.12
assumed*4: E=0.08 assumed

Attachment-13

Estimation of the conditions of structural materials in the Unit-1 containment vessel

1.

OutlineA significant amount of fuel debris is estimated as highly possible to have been relocated in Unit-1 anderoded the concrete surrounded by the RPV pedestal. This document investigates the conditions ofstructural materials in the containment vessel (PCV) in such a situation.

2.

Structural configuration in the Unit-1 containment vessel
Figure 1 shows the structural configuration in the Unit-1 PCV. The RPV loads (dead weights, seismic

loads and other loads) are transferred first to the RPV pedestal made of reinforced concrete andeventually to the base mat made of reinforced concrete via steel inner skirts.

3.

Part of concrete eroded by the fuel debris
The fuel debris is estimated currently (November 2011) to have been relocated to the position shownin Figure 1 and to have eroded the concrete as shown in Figure 2. Drain sump pits were located where
the fuel debris is estimated to have been relocated, and the fuel debris sedimented in the pits was
assumed to have eroded the concrete around. This concrete was not a stiff member of the structure andhad no rebars to bear any loads (it had several exceptional steel bars but they were to prevent cracking).
The concrete eroded by the fuel debris would be difficult to bear loads, but it is of no significancebecause the part currently estimated to have been eroded was not designed originally to bear any loads.
Thus, as long as the fuel debris is assumed to remain where it is currently estimated to be present, it
has no direct impacts on the structural members to support heavy structures in the PCV such as the
RPV and therefore, the structural integrity can be considered to be ensured.

Attachment 13－1

Structures in Unit-1 PCV

34mm

16mm

Biological
生体遮蔽
shieldingRPV

4 openings for CRD
CRD用開口４カ所（68,112,248,292度）°
1000X690, OP13340

OP14660

PCV格納容器赤道
equatorOP12200

Drain
sump(45,225°)
ドレンサンプ（45,225度）

ID50001450X1450X1200

OD7400RPV
pedestal, 1200t
RPVペデスタル
Opening
1200厚 at the bottom
下部に770幅、１９００高の(770
width, 1900
開口あり(190度)
height,190 degrees)

OP10200

Pump
sump(45,225°)
ポンプサンプ（
45,225度）
1400X1700X1230

Fueldebris
燃料デブリ

OP6180150D

OP3570OP2150Base mat at OP-4000

基礎マット下面はOP-4000°°

鋼製内側スカート、D6200
Steel inner skirt,D6200
34t, Manhole at34厚、0,180度にマンホール
0 and 180 degrees

Figure 1

Drywell skirt,D10000
ドライウエルスカート、D10000
15t, Manhole
15厚、マンホールあり

Structures in Unit-1 PCVAttachment 13－2

Pedestal outer surface
Inner skirt
0.65m

Pedestal inner surface

1.45m
2.5m
3.1m
0.75m

3.7m

Cross-sectional configuration
at the erosion depth
Drain sump pit

Slit

0.65m

1.45m
2.5m
3.1m

Cross-sectional configuration
at the erosion depth

0.75m

3.7m

The erosion volume surface is considered to become round-shaped as the sump pit walls
became eroded by the fuel debris sediment in the drain sump pits.
The actual cross-sectional configuration of the erosion volume will be a geometry between
the top and bottom figures.
Figure 2

Schematic of concrete erosion
Attachment 13－3

Attachment-14

Results of in-containment gas composition analysis

1.

OutlineGas samples were collected from the Unit-1 and Unit-2 containment vessels (PCVs) on July 29,
August 9 and September 14, 2011, to measure concentrations of radioactive materials in the PCV(Figure 1, Figure 2). The collected samples were analyzed to obtain gas compositions of hydrogen,carbon monoxide and carbon dioxide. PCV gas samples analyzed were the following nine lots.


Lot Nos. 1 to 3: Unit -1 PCV gas samples (collected on July 29, 2011),



Lot Nos. 4 to 6: Unit -2 PCV gas samples (collected on August 9, 2011), three lots

Lot Nos. 7 to 9: Unit -1 PCV gas samples (collected on September 14, 2011), three lots

three lots

Lot Nos. 1 to No. 3 were collected on July 29, 2011. Lot No. 1 is the gas vial containing the gas
sample collected on July 29, 2011, and Lot Nos. 2 and 3 are two gas vials containing the samples
redistributed on October 3, 2011, from the lot having been stored after collection (collected on July 29,2011).
Lot Nos. 4 to No. 6 were three lots collected on August 9, 2011, from the Unit-2 PCV.
Lot Nos. 7 to No. 9 were three lots collected on September 14, 2011, from the Unit-1 PCV.
Table 1 shows the results of analysis. The results are the gas concentrations in the gas vials.
Hydrogen was detected in both Unit-1 and Unit-2 samples. Carbon monoxide was detected in Unit-2
sample (at the level of the detection limit), but not in Unit-1 sample. Carbon dioxide was detected in both
Unit-1 and Unit-2 samples.

Attachment 14－1

Figure 1

Figure 2

Temporary PCV gas sampling rack (Unit-1, July 2011)

Temporary PCV gas sampling rack (Unit-2 in August 2011 and Unit-1 in September 2011)

Table 1Results of gas analyses (concentrations in vials)
Attachment 14－2

(unit: vol %)Lot No.

Sample

H

CO

CO2

1

Unit-1 (July) ①<0.001<0.01

0.139*１

2

Unit-1 (July) ②<0.001<0.01

0.133*１

3

Unit-1 (July) ③<0.001<0.01

0.112*１

4

Unit-2 (August) ①

0.507

0.014

0.145

5

Unit-2 (August) ②

0.964

0.015

0.143

6

Unit-2 (August) ③<0.001

<0.01

0.145

7

Unit-1 (September) ①

0.14

<0.01

0.114

8

Unit-1 (September) ②

0.092

<0.01

0.189

9

Unit-1 (September) ③

0.072

<0.01

0.124

10

Air (outdoors)<0.001

<0.01

0.074

Information

Detection limit

0.001

0.01

0.01

*1 Information only because of the high dilution in air and the influence of CO2 concentration in the air.
2.

Concentrations of gas compositions in the containment vessel (PCV)

Lot Nos. 1 to No. 3 (collected in July 2011, from Unit-1 PCV) were diluted in the sampling tools with air
into 4.32 times by volume, and then the 6 mL diluted gas was injected with a syringe into vacuumed vials
of 14.1 mL. Before filling, the vials contained the indoor atmosphere (in the analysis room), unlike Lot
Nos. 4 to No. 9. Consequently, the PCV gas volume in the vial was calculated to be 1.4 mL, the rest
being the air. The PCV gas concentrations (concentrations in the non-condensable gas) can be obtainedby Eq. 1 below.

A

14.1C  12.7 B
1.4

Eq. 1

Here, A is the concentration in the non-condensable gas, B is the concentration in air and C is theconcentration in in the vial.

From Lot Nos. 4 to No. 9 (collected in August 2011 from Unit-2 and in September 2011 from Unit-1), thesample gas was directly extracted from the sampling tools, and then the 12.8 ml sample gas was
injected into vacuumed vials of 14.1 mL. Before filling, the vials contained the outdoor atmosphere.
Consequently, the PCV gas concentrations (concentrations in the non-condensable gas) can becalculated by Eq. 2 below.

A

14.1C  1.3B
12.8

Eq. 2

Attachment 14－3

Table 2 shows the gas concentrations derived from the analysis results and corrected for dilution by air.
The concentrations in Lot No. 10 was used as concentrations in the air, as needed for corrections.
Concentrations of hydrogen and carbon monoxide were set at the value of detection limits.
It should be noted that the estimation of the in-PCV gas carbon dioxide concentration was concluded
to be not possible from the Lot Nos. 1 to No. 3 for the following reasons: these samples collected in
Unit-1 (July 2011) were diluted with air in a high dilution ratio and about 90% of the diluted gas samples
were indoor air with higher carbon dioxide contents as compared with the outdoor air. Consequently,
most carbon dioxide detected in Lot Nos. 1 to No. 3 was considered to have originated in the diluted
indoor air; Table 3 shows the results of sensitivity analysis of in-PCV gas carbon dioxide concentrations
derived by Eq. 1 when the in-air carbon dioxide concentration was parametrically changed between
0.038 % and 0.20 %. If the in-air carbon dioxide concentration was low, the in-PCV gas carbon dioxide
concentration should become very high, while if it was high, the in-PCV gas carbon dioxide
concentration should become negative; the in-air carbon dioxide concentration predominates theestimation results of the in-PCV gas carbon dioxide concentration, but the carbon dioxide concentrationin the diluted air in the vial was not accurately known.

Table 2Concentrations in the PCV non-condensable gases (Corrected for dilution by air)(unit: vol %)Lot No.

Sample

H

CO

CO2

1

Unit-1 (July) ①<0.001<0.01

See Table 62Unit-1 (July) ②<0.001<0.01

See Table 6

3Unit-1 (July) ③<0.001<0.01

See Table 6

4Unit-2 (August) ①

0.558

0.014

0.152

5

Unit-2 (August) ②

1.062

0.016

0.150

6

Unit-2 (August) ③<0.001

<0.01

0.152

7

Unit-1 (September) ①

0.154

<0.01

0.118

8

Unit-1 (September) ②

0.101

<0.01

0.201

9

Unit-1 (September) ③

0.079

<0.01

0.129

Table 3

In-PCV gas CO2 concentrations estimated from the in-air CO2 concentration(sensitivity analysis, Sample 1 collected in July 2011)(unit: vol %)Lot No.

Sample1In-air CO2 concentrations assumed
0.038％

0.074％

0.20％

Unit-1 (July) ①

1.055

0.729

－0.4142

Unit-1 (July) ②

0.995

0.668

－0.4753

Unit-1 (July) ③

0.783

0.457

－0.686

Attachment 14－4

3. Situation of core-concrete reactions estimated from CO and CO2 concentrations
Gas concentrations of carbon monoxide and carbon dioxide in the Unit-1 and Unit-2 PCVs were
estimated in a situation if core-concrete reactions had been ongoing when the gas samples had beencollected. The DECOMP, a MAAP built-in sub-program for core-concrete reaction evaluation, was used
to evaluate the core-concrete reactions. Table 4 shows the generation rate and concentration of carbon
monoxide and carbon dioxide gases when the fuel debris having been relocated to the pedestal reactedwith concrete. The non-condensable gas concentrations in the PCV were evaluated in terms of the ratio(CO+CO2)/(N2+CO+CO2). If the core-concrete reactions had been ongoing at the time of gas sampling,
the concentration of CO+CO2 should be more than 10%, but the concentration observed in the gas
composition analysis was significantly lower (Table 1 and Table 2). Gas compositions observed (Table 1
and Table 2) are different from the gas compositions due to core-concrete reactions by computational
analysis (Table 5), too. It is unlikely that the core-concrete reactions are in progress now (November2011).

Table 4

Estimation of the amount of CO and CO2 production and concentration
due to core-concrete reactions (computational analysis) *1
Unit-1

Fraction of fuel relocation to PCV (%)

Amount of production(kmol)(accumulated)

Gas production rate
(Nm3/h)(average) *2
CO+CO2 concentration in
non-condensable gas (%)

100%

Unit-2
57%

57%(via instrumentation(via CRD

line penetration)

penetration)

Steam

75.8

1.18

4.3

H2

37.8

2.06

4.1

CO

5.2

0.04

0.1

CO2

8

0.12

0.46

Steam

15.4

19.8

24.6

H2

7.6

34.8

23.4

CO

1.0

0.8

0.6

CO2

1.6

2.0

2.6

CO+CO2/N2+CO+CO2

8.5

17.7

19.8

*1

Gas production in the pedestal sump pits only.

*2

The accumulated amount divided by the time duration of core-concrete reactions.

Attachment 14－5

Table 5

Gas compositions in core-concrete reactions (DECOMP)
Unit-1100%

Fraction of fuel relocation to PCV (%)-

Unit-2
57%

57%(via instrumentation(via CRD

line penetration)

penetration)

Compositions of

H2

75%

93%

88%

Non-condensable gas

CO

10%

2%

2%

generated

CO2

16%

5%

10%

Supplement (1) Discharge of carbon dioxide dissolved in the injected water to PCV and RPV
The water that had been injected to the reactor at the time of gas sampling (July to September2011) is considered to have contained a certain degree of carbon dioxide (free carbonic acid). The
properties of the water can be considered to be close to those of surface water or dam water. Table
6 shows the estimated carbon dioxide concentration discharged in the PCV, when the free carbonicacid concentration was 2 to 10 mg/L (Water Statics 2000, MLFW), which is close to the value in thesurface water or dam water.
Depending on its concentration in the injection water, the free carbonic acid dissolved in theinjection water can influence the carbon dioxide concentration of about 0.01% to 0.16%. The carbon
dioxide concentration observed in the PCV gas samples may have been significantly influenced by
the free carbonic acid dissolved in the injection water.

Table 6

CO2 amount contained in the injection water and estimated CO2 concentrationin the PCV non-condensable gases
3

Amount of water injection [m /h]

Unit-1

Unit-2

4

4

Amount of N2 gas injection [Nm3/h]

28

CO2 concentration in water [mg/L]
Estimated CO2 concentration(in non-condensable gases) [%]
-

4

4
13

2

10

2

10

0.01

0.07

0.03

0.16

Supplement (2) Thermal decomposition of polymer compounds used for electric cables and others
Polymer compounds in the PCV such as electric cable covering materials generate carbon
monoxide and carbon dioxide when heated up. But the current PCV temperatures have been below
the level of polymer thermal decompositions (for example, about 200 deg C for vinyl chloride resins)from the time of sample collection till now (November 2011). Therefore, it is unlikely that carbon
monoxide or carbon dioxide were and are now being generated during this time period.

Attachment 14－6

Attachment-15

Gas residues from the early phase of core-concrete reactions

1.

Summary
A large amount of steam was being generated in the reactor after the accident by water injection forfuel cooling. Gases in the containment vessel (PCV) could leak out at the time of interest, and therefore
non-condensable gases generated in the early phase of the accident have unlikely remained till now(November 2011) in significant amounts. On the other hand, a small amount of rare gas (Kr-85) was
detected in the gas samples of Unit-2, indicating that part of the non-condensable gases generated in
the early phase of the accident could have remained locally.

2.

PCV gas replacement due to steam production
A large amount of steam was being generated in the reactor after the accident by water injection forfuel cooling. Figure 2 shows the amount of water injection which could remove the fuel decay heat of
Unit-2 by the latent heat for boiling (equivalent to the decay heat). If the amount of steam generated isassumed to be the amount of water injected equivalent to the decay heat, the total amount of steam
generated between April 1, 2011 and June 28, 2011, when nitrogen gas filling started is about 6300 t,
which is equivalent to about 1.1x107 m3 of steam at atmospherical pressure and 100 deg C. Assuming
the Unit-2 PCV open volume as about 3500 m3 (with no water accumulation), the total amount of steam
generated is about 2400 times the PCV open volume. As the gases in the reactor pressure vessel (RPV)and PCV could leak out at the time of interest, the PCV gases could have been replaced about 2400
times. Consequently, non-condensable gases can be considered unlikely to have remained in asignificant amount in the PCV.

Attachment 15－1

10.009.00

崩壊熱相当注水量[t/h]

8.00
7.00
6.00
5.00
4.00
3.00
2.00
1.00
0.00
March,1st
3月1日 March,16th
3月16日 March,31th
3月31日 April,15th
4月15日

April,30th
4月30日

May,15th
5月15日

May,30th
5月30日

June,14th
6月14日

June,29th
6月29日

July,14th
7月14日

2011

Figure 1

3.

Amount of water injection equivalent to decay heat of Unit-2

Residue fractions in the Unit-2 PCV gases
A small amount of rare gases (Kr-85 and others) was detected in August 2011 upon PCV gas sampling

and gamma analysis of sampled gases of Unit-2. Kr-85 has a long half-life and its yield fromspontaneous fissions is negligibly small. Therefore, the detected Kr-85 is considered to have originatedand been accumulated during normal plant operations. The residue fraction of Kr-85 can be obtained as
the ratio between the initial inventory and in-PCV inventory (to be estimated from the detectedconcentration of Kr-85).

AKr85 （１－）
 VPCVM Kr85Eq. 1

α: residue fraction, β: steam fraction, VPCV: PCV volume
AKr85: Kr-85 concentration in non-condensable gases
Mkr85: Kr-85 initial inventory
If this residue fraction is applied to other non-condensable gas components, the initial inventories ofeach gas component can be inversely calculated.

M GAS 



AGAS 
（１－） VPCVAGAS
M Kr85
AKr85

Eq. 2

AGAS: Gas concentrations in the non-condensable gases,
MGAS: Gas initial inventories
Attachment 15－2

The Kr-85 initial inventory was set as the inventory 150 days after reactor shutdown, by taking theattenuation by decay into account. If the carbon dioxide detected in the samples collected in August
2011 (see Attachment-14) is assumed to be the residue of carbon dioxide which had been generated at
an early phase of the accident only, the carbon dioxide initial inventory at the early phase of the accident
can be estimated by Eq. 2 and the results are shown in Table 1.

Table 1

Carbon dioxide initial inventory at Unit-2 estimated from Kr-85 residue ratio

Kr-85 concentration in the non-condensable gases [Bq/cm3]

74.5

Kr-85 initial inventory (150 days after reactor shutdown by ORIGEN [Bq]2.20E+16

CO2 concentration in the non-condensable gases [％]

0.151
3

CO2 concentration in the non-condensable gases [mol/cm ]6.76E-08

CO2 initial inventory [mol]2.00E+07

Attachment 15－3

4.

Residue ratios in the Unit-1 containment vessel
It is considered to be unlikely that the non-condensable gases generated at an early phase of theaccident remain in the PCV. But, the residue ratios at Unit-1 were also examined in the following threecases, by considering that a small amount of rare gas was detected at Unit-2. At Unit-1, too, the gamma
ray nuclei analysis (using germanium detectors) recorded signals which could have come from Kr-85 of
the sample collected in July 2011, although the rare gases were as a whole below the detection limits.
Case 1: The same Kr-85 residue ratio as that of Unit-2 was assumed.
Case 2: The estimated amount of Kr-85 in the July 2011 sample (below detection limit) was used toestimate the residue ratio.
Case 3: The residue ratio was estimated by exponentially relating the ratio (of the Kr-85 amount)between the July sample and the September sample.

4.1.

Case 1

Gas leak paths from the PCV are unknown and nitrogen gas filling started in early April at Unit-1. The
residue ratio will be different in Unit-1 and Unit-2. It will overestimate the initial inventories to assume thesame residue fraction in Unit-1 as that in Unit-2. If the residue fraction were assumed to be equal, and if
the carbon dioxide detected came from only the carbon dioxide having been generated in the earlyphase of the accident, the initial inventories will be such as given in Table 2.

Table 2

(Case 1) CO2 initial inventory of Unit-1 estimated from the residue ratio of Unit-2
Case 1

July 2011 *September 2011

CO2 concentration in the non-condensable gases [％]6.18E-01

1.49E-01

CO2 concentration in the non-condensable gases [mol/cm3]

2.76E-07

6.66E-08

CO2 initial inventory [mol]8.15E+07

1.97E+07* Information only due to high dilution with air and the in-air CO2 concentration dominates the results.
4.2.

Case 2At Unit-1, the gamma ray nuclei analysis (using germanium detectors) recorded energy peaks which
could have come from Kr-85 from the sample collected in July 2011, although the amount was below thedetection limit. If this peak were really from Kr-85, the amount should have been 4.67x102 Bq/mL (thedetection limit was about 5.31x102 Bq/mL*). Table 3 shows the initial inventories derived by Eq. 2 using
this measured value. Numerically, the results are in the same order with the results in Case 1, but thevalue from the Unit-1 July sample should be regarded as information only, because the sample wasdiluted with air in a high dilution ratio and the in-air CO2 concentration dominates the results.
*The actual detection limit is about 1.23x102 Bq/mL. The value of 5.31x102 Bq/mL is the
non-condensable gas concentration equivalent to the detection limit when the effect of dilution by 4.32
times is taken into account.

Attachment 15－4

Table 3(Case 2) CO2 initial inventory of Unit-1 estimated from the Kr-85 residue ratio
July 2011 *CO2 concentration in the non-condensable gases [％]6.18E-01

CO2 concentration in the non-condensable gases [mol/cm3]

2.76E-07

CO2 initial inventory [mol]1.01E+07* Information only due to high dilution with air and the in-air CO2 concentration dominates the results.
4.3.

Case 3

If the non-condensable gas concentrations in the PCV can be assumed to decrease exponentially, theresidue fraction can be estimated from the ratio of concentrations in the samples collected in July 2011
and in September 2011. Table 4 shows the residue fractions and carbon dioxide initial inventory thus
estimated. By using the residue fractions, the amount of Kr-85 can be inversely calculated at any time
points. The results are shown in Table 5. The amount of Kr-85 residue as of July 2011, the timing ofsample collection, far exceeds the detection limit (about 5.31x102 Bq/mL*), which is inconsistent with theresult of gamma ray nuclei analysis (using germanium detectors). Case 3 can be concluded as “notapplicable.”*The actual detection limit is about 1.23x102 Bq/mL. The value of 5.31x102 Bq/mL is the
non-condensable gas concentration equivalent to the detection limit when the effect of dilution by 4.32
times is taken into account.

Table 4CO2 initial inventory estimated from the concentration ratio in July and September 2011
Unit-1 Average CO2 concentration in July 2011 [％] *0.618

Unit-1 Average CO2 concentration in September 2011 [％]0.149

Attenuation ratio (September/July)

2.42E-01

Attenuation rate per day

9.70E-01

Initial CO2 Inventory [mol]2.84E+06* Information only due to high dilution with air and the in-air CO2 concentration dominates the results.
Table 5

Kr-85 residue amount estimated from Case 3 residue fraction
July 2011

Kr-85 initial inventory
Inventory in PCV

September 2011
1.70E+16

2.47E+14

5.97E+13

3

PCV open volume [m ]

2800

2800

Steam fraction in PCV

0.65

0.46

8.82E+04

2.13E+04

2.52E+05

3.95E+04

Estimated radioactivity concentration in
PCV [Bq/cm3]
Estimated radioactivity concentration in
non-condensable gases [Bq/cm3]

Attachment 15－5

5.

Gas residues from the early phase of the accident estimated from the gas residue fractions
Table 6 gives the initial inventories of carbon monoxide and carbon dioxide estimated by DECOMP,

the MAAP built-in sub-program for core-concrete reaction evaluation. They are at most 1.4x104 mol,about 3 orders of magnitude less than the values given in Table 1 to Table 3, which are in the order of 10
to the 7th power. In estimating residue fractions, the initial inventories tend to be overestimated if anadditional amount flows in from outside during the process. It is estimated, therefore, that the carbon
dioxide detected in the sampled PCV gases had flowed in by unknown reasons to the PCV from outside.
Even if large scale core-concrete reactions had occurred at an early phase of the accident, hardly any of
the gases generated at the time would have remained till now and the gases that flowed in during theprocess would be currently dominant.

Table 6

Amount of CO and CO2 at the time of core-concrete reactions (analysis by DECOMP)
Unit-1

Unit-2
57%

Fraction of fuel relocation to PCV (%)

100%

Instrumentationline penetration

57%
CRD penetration

Steam

75.8

1.18

4.3

Integrated amount of gas generated

H2

37.8

2.06

4.1

(by analysis) (kmol)

CO

5.2

0.04

0.1

CO2

8

0.12

0.46

CO+ CO2

13.2

0.16

0.56

Note: Gases generated in the pedestal sump pits only

Attachment 15－6

Attachment-16
Paint stripping-off incidents of Unit-2 reactor building ceiling cranes
1.

Outline of the incident
In a photo taken on September 17, 2011, a white gas, probably steam, was observed tobe blowing out from immediately above the reactor. About one month later, on October 20,
2011, no gas blowout was observed in the photo but some stripping-off was recognized onpart of the ceiling crane coatings. This stripping-off is considered to have occurred after thesteam blowout ceased.
2.

Stripping-off mechanism
Defects of coating such as stripping-off or cracking are generally considered to be

attributed to the internal stresses generated in the coatings. When the internal stresses
exceed the coating adhesion strengths, the coatings strip off. Coatings bear contraction
stresses ever since they were constructed at the beginning. The stresses repeat ups anddowns in response to ambient temperature changes. It is known that the internal stresses
gradually decrease with time in a higher humidity environment. The stresses mitigated bymoisture absorption increase again when dehydrated [1].
Therefore, the mechanism of ceiling crane coating stripping-off at Unit-2 can beconsidered as follows:① The coatings were deteriorated by the heat and steam at the time of the accident andtheir adhesion strengths decreased.
② Water injection to the reactor for fuel cooling suppressed steam production and theambient temperatures decreased. As the result, the coatings were dehydrated, and theinternal stresses recovered to higher values.
③ The coatings stripped off when the recovered internal stresses exceeded the lowered
adhesive strengths of the coatings.
Reference[1] Kozo Sato, Adhesion of coatings, its theory of, and commentary to, the mechanism,
RIKO Publishing Corporation, 1981 (in Japanese).
End

Attachment 16-1