原子炉圧力(MPa[abs]) Reactor pressure (MPa[abs]) Primary containment vessel pressure (MPa[abs])

(1)

MAAP Code-based Analysis

of the Development of the Events

at the Fukushima Daiichi Nuclear Power Station

March 12, 2012

Tokyo Electric Power Company

(2)

1 Contents of This Report

1. Introduction 2. Unit 1

1. Issues to be solved regarding the previously published analysis results of Unit 1 2. Main presumptions made for the analysis of Unit 1

3. Analysis results of Unit 1 3. Unit 2

3. Analysis results of Unit 2 4. Unit 3

3. Analysis results of Unit 3 5. Conclusions

(3)

2 1. Introduction

 On May 23, 2011, we conducted a MAAP-code-based plant-state analysis based on information that had been obtained on the state of the plant and how the plant had been operated immediately after the earthquake, and made the analysis results public.

 Because there were inconsistencies at that point between the observation data and the analysis results, we have continued conducting hearings from operators and on-site surveys since the publication in order to obtain more detailed information on the plant.

 In parallel with the hearings and surveys, we have developed, from data taken from the actual units in the plant and the MAAP analysis results, plant-state presumptions that allow the course of the accidents to be rationally accounted for.

 The present analysis was conducted with the analysis conditions set in such a way that the behavior of the plant immediately after the accidents can be reproduced as accurately as possible based on the information of the plant obtained to date and our plant-state

presumptions.

(4)

3 2.1. Issues to be solved regarding the MAAP analysis results of Unit 1 published in

May 2011

0 2 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

日時

原子炉圧力(MPa[abs])

RPV圧力（解析）

実機計測値（ A系）

実機計測値（ B系）

IC起動による圧力低下

RPV破損（約15時間後）

原子炉建屋爆発（約25時間後）

炉心損傷開始（約4時間後）

課題①

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6

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 日時

原子炉格納容器圧力(MPa[abs])

D/W圧力（解析）

S/C圧力（解析）

実機計測値（ D/W）

実機計測値（ S/C）

RPV破損(約15時間後) S/Cベント

格納容器漏えいを仮定(約18時間後以降)

格納容器漏えい拡大を仮定(約50時間後以降)

炉心損傷開始（約4時間後）

原子炉建屋爆発（約25時間後）

課題②

Reactor-pressure values from the analysis are inconsistent with the measured

reactor-pressure values

Reactor pressure (MPa[abs])

RPV pressure (analysis)

Measurement from actual unit (A system) Measurement from actual unit (B system)

Pressure decrease due to IC startup

RPV damage (approx. 15 hours later)

Reactor building explosion (approx. 25 hours later) Start of core damage (approx. 4 hours later)

Issue 1

Time and date

Time and date Primary containment vessel pressure (MPa[abs])

D/W pressure (analysis) S/C pressure (analysis)

Measurement from actual unit (D/W) Measurement from actual unit (S/C)

S/C vent Presumed leakage from the PCV

(from approx. 18 hours later)

Presumed buildup of the leakage from the PCV (from approx. 50 hours later)

Start of core damage (approx. 4 hours later)

Issue 2

Primary containment-vessel (PCV) pressure values from the analysis are inconsistent with the measured PCV pressure values.

Reactor building explosion (approx. 25 hours later)

(5)

4 2.2. Main presumptions made for the analysis of Unit 1

 It was presumed that the emergency condenser (IC) had not operated since the loss of all AC power due to the tsunami.

 It was presumed that a leakage from the gaseous phase section of the reactor pressure vessel (RPV) had occurred when the maximum fuel clad temperature of 727C (1000 K) had been reached and when the temperature of the gas in the reactor had reached 450C respectively.

 To reproduce the behavior of the PCV pressure, it was presumed that a gaseous phase leakage from the PCV had occurred approximately 12 hours, 50 hours and 70 hours after the earthquake.

< Main presumptions >

Issues 1 and 2

(6)

5 2.2. Gas leakage locations (1)

If the fuel is exposed and the temperature of the core becomes high, the nuclear instrumentation piping in the core may be damaged.

If the dry tubes of the SRM/IRM and TIP, which have a mechanism for extraction to outside the reactor, are damaged, steam in the gaseous section in the reactor may leak to D/W.

In this analysis, it was presumed that a leakage had occurred when the maximum temperature of

the fuel clad had reached 727 C (1000 K).

(Cross section of leakage hole is presumed as 0.00014 m²)

索引装置 PCV内

SRM/IRM 駆動装置

中性子束計装ハウジング

継ぎ手部

PCV外 SRM/IRM

ドライチューブ TIPドライチューブ

索引装置 PCV内

SRM/IRM 駆動装置

中性子束計装ハウジング

継ぎ手部

PCV外 SRM/IRM

ドライチューブ TIPドライチューブ

Index equipment Inside the PCV

SRM/IRM drive

Neutron flux

instrumentation housing

Joint section

SRM/IRM dry tube

Outside the PCV TIP dry tube

(7)

6 2.2. Gas leakage locations (2)

The gaskets used in the flange sections of such parts as the SRV pipe beds of the main steam piping may lose their sealing capability if the fuel is exposed; the temperature of the core becomes high, and the temperatures of the gases (steam and hydrogen) in the core become high.

In particular, the temperature the expansion

graphite gaskets can withstand is approx. 450 ℃.

In this analysis, it was presumed that a leakage had occurred when the gas temperature in the core had reached 450C.

(Cross section of leakage hole is presumed as 0.00136 m²)

SRV main body

Steam

Inlet flange

Outlet flange

(8)

7

-10 -8 -6 -4 -2 0 2 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

日時

原子炉水位（ m ）

ダウンカマ水位シュラウド内水位

実測値(原子炉水位（燃料域)(A）) 実測値(原子炉水位（燃料域)(B）) TAF到達

3月11日18時10分頃 BAF到達

3月11日19時40分頃

注水開始

TAF

BAF

2.3. Overview of the analysis results of Unit 1 (reactor water level)

Time and date

Reactor water level (m)

Downcomer water level Water level inside the shroud

Measured value (reactor water level (Fuel Region A)) Measured value (reactor water level (Fuel Region B)) TAF reached

(around 18:10, March 11) BAF reached

(around 19:40, March 11)

Water injection started

(9)

8

0 2 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

日時

原子炉圧力(MP a[ ab s] )

原子炉圧力

実測値（A系原子炉圧力）

実測値（B系原子炉圧力）

IC起動による圧力低下

RPV破損

主蒸気配管フランジからの気相漏えい

炉内核計装からの気相漏えい

溶融燃料の下部プレナムへの落下による圧力上昇

2.3. Overview of the analysis results of Unit 1 (reactor pressure)

Time and date

Reactor pressure

Measured value (reactor pressure, A system) Measured value (reactor pressure, B system)

Pressure drop due to the activation of IC

RPV damage Gaseous phase leakage

from flange sections of the main steam piping Gaseous phase leakage

from the within-reactor nuclear measurement instrumentation piping

Pressure rise due to the dropping of molten fuel to the lower part of the plenum

The behavior of the reactor pressure is reproduced by presuming that there were gaseous phase leakages from the reactor pressure vessel.

A sharp pressure rise is seen in the analysis result, but this is attributable to the

characteristics of MAAP models, and it is most likely that this did not occur in reality.

(10)

9

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6

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

日時

格納容器圧力( Mp a[ ab s] )

D/W圧力 S/C圧力

実測値(D/W圧力) 実測値(S/C圧力)

RPV破損

主蒸気配管フランジからの気相漏えい

炉内核計装からの気相漏えい

溶融燃料の下部プレナムへの落下による圧力上昇

S/Cベント

格納容器漏えいを仮定

2.3. Overview of the analysis results of Unit 1 (Primary containment vessel pressure)

The behavior of the primary containment vessel pressure is reproduced by presuming that there were gaseous-phase leakages from the reactor pressure vessel.

Time and date

Primary containment vessel pressure (Mpa[abs])

D/W pressure S/C pressure

Measured value (D/W pressure) Measured value (S/C pressure)

RPV damage

Gaseous-phase leakage

from flange sections of the main steam piping Gaseous-phase leakage

from the within-reactor nuclear measurement instrumentation piping

Pressure rise due to the dropping of molten fuel to the lower part of the plenum

S/C vent

Presumed leakage from the PCV

(11)

10 2.3. Summary of the analysis results of Unit 1

< Summary of the analysis results >

- Time at which the exposure of the core started:

around 18:10, March 11 (approx. 3 hours after the earthquake) - Time at which the core started being damaged:

around 18:50, March 11 (approx. 4 hours after the earthquake) - Time at which the RPV was damaged:

around 1:50, March 12 (approx. 11 hours after the earthquake)

 As a result of the adoption of the presumption of gaseous phase leakage, we succeeded in reproducing the reactor-pressure behavior (Issue 1) and PCV pressure behavior (Issue 2) fairly accurately.

 Because of the presumption that the IC had not operated since the loss of all AC power due to the tsunami, the damage of the core and the RPV occurred relatively early in the analysis.

 The time at which the RPV damage occurred is about 4 hours earlier. This is a result of the unrealistic pressure behavior due to the characteristics of the MAAP code, and we think that further enhancement of the analysis code should be made.

(12)

11 3.1. Issues to be solved regarding the MAAP analysis results of Unit 2

published in May 2011

-2 0 2 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

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00 日時

原子炉圧力 (MPa[abs])

RPV圧力(解析）

実機計測値

RCIC起動 RCIC停止

SRV開

RPV破損(約109時間後) 炉心損傷開始(約77時間後)

課題③

-0.2 0 0.2 0.4 0.6 0.8 1

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 日時

原子炉格納容器圧力 (MPa[abs])

D/W圧力(解析）

S/C圧力(解析) 実機計測値(D/W) 実機計測値(S/C)

D/Wに漏えいを仮定 (約21時間後）

SRV開

RCIC停止

RCIC起動

炉心損傷開始(約77時間後)

RPV破損(約109時間後)

課題④ 課題⑤

Reactor pressure values from the analysis are inconsistent with the measured-reactor pressure values

If it is presumed that there were leakages, it will not be possible to reproduce measured values that allow a high primary-containment-vessel pressure level to be maintained.

PCV pressure values from the analysis are inconsistent with the measured PCV pressure values unless it is presumed that there were leakages from the primary containment vessel.

Time and date

RPV pressure (analysis) Measurement from actual unit

RCIC startup RCIC shutdown

SRV opened

RPV damage (approx. 109 hours later) Start of core damage

(approx. 77 hours later)

Issue 3

Time and date

Primary containment vessel pressure(MPa[abs])

Presumed leakage (D/W) (approx. 21 hours later)

SRV opened

RCIC shutdown

RCIC startup

Issue 4 Issue 5

Start of core damage (approx.

77 hours later)

(13)

published in December 2011

-0.2 0 0.2 0.4 0.6 0.8 1

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

時刻

格納容器圧力 (MPa[abs])

実機計測値(D/W) 実機計測値(S/C) D/W圧力(解析）

S/C圧力(解析)

RCIC停止判断 SRV開

Issue 8

An analysis result was obtained that allowed high PCV pressure level to be maintained, but the analysis values did not quantitatively

reproduce the measured values (because of the small volume of the hydrogen generated).

Time

Primary-containment-vessel pressure (MPa[abs])

Measurement from actual unit (D/W) Measurement from actual unit (S/C) D/W pressure (analysis)

S/C pressure (analysis)

RCIC shutdown decided SRV opened

(14)

and a summary of the analysis results of Unit 2

< Main presumptions made for the analysis of Unit 2 >

 It was presumed that the steam to the reactor core isolation cooling system (RCIC) turbine had been a two-phase flow condition with a level of energy equivalent to the decay heat, so that a pressure behavior can be reproduced, which is characterized by the staying of the pressure at a level below the SRV activation pressure setting.

 It was presumed that gradual ingress of water into the torus room after the arrival of the tsunami had caused the heat in the PCV to be transferred from the S/C boundary to outside the PCV.

 It was presumed that the amount of water injected into the reactor using fire engines had been an amount that had raised the reactor water level to a level below the fuel region taking into consideration the volume of hydrogen generated.

 To reproduce the behavior of the PCV pressure, it was presumed that a gaseous-phase leakage from the PCV had occurred approximately 89 h after the earthquake.

Issues 4 and 5

Issue 8 Issue 3

(15)

14

-10 -8 -6 -4 -2 0 2 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

日時

原子炉水位 (m)

実機計測値（燃料域A)

水位補正値

TAF

BAF

3.2. Operation states of the RCIC in Unit 2

基準面器

差圧計Ｈｓ

Ｈｒ

炉側配管基準面器

側配管

The measured water level coincides with the water level of the reference plane vessel (highest value on the water gauge) after corrections.* It is most likely that the actual water level was higher, and it is considered that steam mixed with water was flowing into the RCIC turbine (in the form of a two-phase flow).

(If there is a power source, the injection of water is stopped at a water level that is lower than the power source (L8). )

*: Corrections based on the reactor pressure and PCV temperature

Time and date

Measurement from actual unit (Fuel Region A)

Water level after correction

Reference plane vessel

Differential pressure

gauge

Piping at reactor side Piping at

reference plane vessel side

(16)

15 3.2. State of the primary containment vessel (S/C) in Unit 2

It was presumed that ingress of water into the torus room, in which the S/C is installed, had occurred and that a heat transfer path had been established, which had allowed the heat transferred to the S/C to be transferred to the water in the torus room via the S/C wall.

The state of the torus room for Unit 2 has not been confirmed, but it has been confirmed that the S/C for Unit 4 was submerged.

Photograph of the area directly below the catwalk in the torus room for Unit 4 (taken from the catwalk)

Water surface

(17)

16 3.3. Overview of the analysis results of Unit 2 (reactor water level)

-10 -8 -6 -4 -2 0 2 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

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00

日時

原子炉水位

(m)

実機計測値（燃料域A）

シュラウド内水位（解析）

ダウンカマ水位（解析）

補正後の水位 RCIC機能低下（仮定）

SRV開

海水注水開始

TAF

BAF RCIC手動起動

TAF到達

BAF到達

Time and date

Measurement from actual unit (Fuel Region A) Water level inside the shroud (analysis) Downcomer water level (analysis) Water level after correction (Presumed) deterioration of the RCIC function

SRV opened

Sea water injection started RCIC manual startup

TAF reached

BAF was reached.

(18)

17

0 2 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

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00

日時

原子炉圧力

(MPa[abs])

実機計測値 RPV圧力(解析）

RCIC手動起動

RCIC機能低下（仮定）

SRV開

計装バッテリ枯渇に伴うハンチング RCICからの注水により

炉心部のボイド率が低下し、原子炉圧力が低下

3.3. Overview of the analysis results of Unit 2 (reactor pressure)

Time and date

Measurement from actual unit RPV pressure (analysis) RCIC manual startup

(Presumed) deterioration of the RCIC function

SRV opened

Hunting due to the depletion of the instrumentation battery The water injection from RCIC reduced the proportion of the voids in the core, which reduced the reactor pressure.

The behavior of the reactor pressure is reproduced by making presumptions about the operation of the RCIC.

(19)

18

0 0.2 0.4 0.6 0.8 1

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

日時

格納容器圧力

(MPa[abs])

実機計測値(D/W) 実機計測値(S/C) D/W圧力(解析）

S/C圧力(解析)

トーラス室に浸入した水によるS/Cの除熱開始

（仮定）

SRV開

D/W気相部から漏えい（仮定）

3.3. Overview of the analysis results of Unit 2 (primary containment vessel pressure)

The behavior of the PCV pressure is reproduced by taking into consideration the heat removal from the external

wall of S/C. PCV pressure characterized by

rise due to the generation of hydrogen and staying at a high level are reproduced.

Time and date

Measurement from actual unit (D/W) Measurement from actual unit (S/C) D/W pressure (analysis)

S/C pressure (analysis)

(Presumed) start of heat removal from S/C by the water that has entered the torus room

SRV opened

(Presumed) leakage from the gaseous phase part of D/W

(20)

 We succeeded in reproducing the reactor-pressure behavior (Issue 3) fairly accurately by making presumptions about the operating conditions of the RCIC.

 We succeeded in reproducing the PCV pressure behavior (Issue 4) fairly accurately and reproducing the PCV pressure behavior characterized by the staying of the pressure at a high level (Issue 5) by making presumptions about the ingress of water into the torus room.

 We succeeded in quantitatively reproducing the pressure values during the period in which the PCV pressure stays at a high level (Issue 8) by adjusting the water volume injected using fire engines.

 The reactor-water level drop due to the RCIC shutdown led to damaging of the core, but the RPV damage did not occur.

(According to our total evaluation based on the plant data obtained to date and other information, it is most likely that RPV damage had occurred.)

(21)

published in May 2011

課題⑥

-2 0 2 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

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00 日時

RPV圧力(解析）

実機計測値 HPCI停止

RCIC停止

SRV1弁開 HPCI起動

RPV破損(約66時間後) 炉心損傷開始(約42時間後)

原子炉建屋爆発(約68時間後)

-0.2 0.0 0.2 0.4 0.6 0.8 1.0

3/11 12:00

3/12 0:00

3/12 12:00

3/13 0:00

3/13 12:00

3/14 0:00

3/14 12:00

3/15 0:00

3/15 12:00

3/16 0:00

3/16 12:00

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00 日時

原子炉格納容器圧力 (MPa[abs])

D/W圧力(解析）

S/C圧力(解析) 実機計測値(D/W) 実機計測値(S/C) S/Cベント

S/Cベント

S/Cベント (仮定) SRV1弁開

計装DS/

ハンチング S/Cベント

S/Cベント炉心損傷開始(約42時間後)

原子炉建屋爆発(約68時間後)

RPV破損(約66時間後)

課題⑦

Reactor pressure values from the analysis

are inconsistent with the measured reactor pressure values.

(The trend can be reproduced if it is presumed that leakages occurred from the piping of the high-pressure core injection system (HPCI)).

There are periods in which measured PCV pressure values are higher than the

corresponding analysis values and periods in which the trend in pressure behavior differs between measured values (falling) and analysis values (rising).

Issue 6

RPV pressure (analysis) Measurement from actual unit HPCI shutdown

RCIC shutdown

SRV1 valve opened

RPV damage (approx. 66 hours later) Start of core damage

(approx. 42 hours later).

Reactor building explosion (approx. 68 h later)

Time and date

Measurement from actual unit (D/W) Measurement from actual unit (S/C) S/C vent

S/C vent

(Presumed) S/C vent SRV1 valve

opened

Instrumentation DS/

hunting S/C vent

S/C vent Start of core damage

(approx. 42 hours later).

Reactor building explosion (approx. 68 hours later)

Issue 7

HPCI startup

(22)

published in December 2011

-2 0 2 4 6 8 10

3/11 12:00 3/11 18:00 3/12 0:00 3/12 6:00 3/12 12:00 3/12 18:00 3/13 0:00 3/13 6:00

日時

RPV圧力(解析）

実機計測値

HPCI停止 RCIC停止

HPCI起動

Issue 9

The reactor pressure decreased monotonically in the analysis, whereas the reactor-pressure measurement shows that the pressure decreased stepwise.

Time and date

RPV pressure (analysis) Measurement from actual unit

HPCI shutdown RCIC shutdown

HPCI startup

(23)

< Main presumptions >

 It was presumed that the RCIC and HPCI had been operating continuously with flow rate adjustments made to keep the reactor water level within the appropriate range (in order to prevent repeated starts and stops due to excessive water level rises and falls).

 A realistic decay heat value was adopted, which reflected the fuel loading history.

 It was presumed that operators had been spraying water into S/C using a diesel-engine- driven firefighting pump.

 It was presumed that the amount of water injected into the reactor using fire engines had been an amount that had keeps the reactor water level to a level below the fuel region.

Issues 6 and 9 Issue 7

Issue 7

(24)

23

復水貯蔵タンク

原子炉圧力容器

MO 主蒸気管

MO

FIC MO

流量制御

タービンタービン

MO

水源切替ライン

MO

圧力抑制室

MO HO HO

AO

給水系

タービン止め弁

加減弁

格納容器

MO

最小流量ﾊﾞｲﾊﾟｽ弁

ミニマムフローライン

注入ライン蒸気管

MO

テストライン

MO

テストバイパス弁

4.2. Operations performed by operators during operation of the HPCI in Unit 3

From operators’ testimonies, it is known that they were adjusting the HPCI flow rate using the test line while confirming the reactor water level. They fully closed the minimum flow line to prevent the S/C water level from rising.

It is also known that they were spraying water into S/C at that time using a diesel-engine-driven firefighting pump.

Condensate storage tank

Reactor pressure vessel

Main steam line

Flow rate control

Water source switching line Suppression

chamber

Water supply system

Turbine stop valve Governor valve

Primary containment vessel

Minimum flow rate bypass valve

Minimum flow line

Injection line Steam line

Test line

Test bypass valve

Turbine

(25)

24

-4 -2 0 2 4 6 8 10

3/11 12:00

3/12 0:00

3/12 12:00

3/13 0:00

3/13 12:00 日時

原子炉水位 (m)

シュラウド内水位(解析）

ダウンカマ水位(解析) 実機計測値

RCIC停止

HPCI起動

4.2. Operation status of the HPCI in Unit 3

It was presumed that the water level had been rapidly restored immediately after the the HPCI startup and that after the restoration of the water level, the HPCI had been operating continuously with flow rate adjustments to

prevent excessive water level rises.

The reactor pressure dropped after the HPCI startup, but it did not drop

monotonically. There is a period during which the pressure drop rate is lower.

0 2 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 1

日時

HPCI停止 RCIC停止

SRV開 HPCI起動

May be attributable to the change in the operation status

of the HPCI.

Time and date

HPCI shutdown RCIC shutdown

SRV opened HPCI

startup

Time and date

Water level inside the shroud (analysis) Downcomer water level (analysis) Measurement from actual unit

RCIC shutdown

HPCI startup

(26)

25 4.3. Overview of the analysis results of Unit 3 (reactor water level)

-10 -8 -6 -4 -2 0 2 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

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00

日時

原子炉水位

(m)

シュラウド内水位（解析）

ダウンカマ水位（解析）

実機計測値

TAF

BAF RCIC停止

SRV開 HPCI停止

淡水注入開始 HPCI起動

海水注入開始

海水注水停止

海水注水再開

建屋爆発海水注水停止

海水注入再開 TAF到達

BAF到達

Time and date

Water level inside the shroud (analysis) Downcomer water level (analysis) Measurement from actual unit RCIC shutdown

SRV opened HPCI shutdown

Freshwater injection started HPCI

startup

Seawater injection started.

Seawater injection stopped

Seawater injection resumed

Reactor building explosion Seawater injection stopped

Seawater injection resumed.

TAF reached

BAF reached

(27)

26 4.3. Overview of the analysis results of Unit 3 (reactor pressure)

0 2 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

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00

日時原子炉圧力

(MPa[abs])

RPV圧力（解析）

実機計測値 HPCI停止

RCIC停止

SRV開 HPCI起動

The reactor pressure drops and the temporary decrease in the pressure drop rate are reproduced from the presumptions of the HPIC operation status.

Time and date

RPV pressure (analysis) Measurement from actual unit HPCI shutdown

RCIC shutdown

SRV opened HPCI

startup

(28)

27

0.0 0.2 0.4 0.6 0.8 1.0

3/11 12:00

3/12 0:00

3/12 12:00

3/13 0:00

3/13 12:00

3/14 0:00

3/14 12:00

3/15 0:00

3/15 12:00

3/16 0:00

3/16 12:00

3/17 0:00

3/17 12:00

3/18 0:00

3/18 12:00

日時

格納容器圧力

(MPa[abs])

D/W圧力(解析）

S/C圧力(解析) 実機計測値(D/W) 実機計測値(S/C)

S/Cベント開 SRV開

S/Cスプレイ開始 S/Cスプレイ停止

S/Cベント開

S/Cベント閉

S/Cベント閉（仮定）

S/Cベント開

S/Cベント開（仮定）

S/Cベント開 S/Cスプレイ開始

D/Wスプレイ開始

S/Cスプレイ停止 D/Wスプレイ停止

4.3. Overview of the analysis results of Unit 3 (primary-containment-vessel pressure)

It was not possible to reproduce the deviations from the PCV pressure measurements.

The magnitudes of the deviations were reduced by the spraying of water into S/C.

Time and date

SRV opened Spraying of water into S/C

started.

Spraying of water into S/C stopped.

S/C vent opened

S/C vent closed

(Presumed) S/C vent closed S/C vent

opened

S/C vent opened

(Presumed) S/C vent closed (Presumed) S/C vent opened

(Presumed) S/C vent closed S/C vent opened Spraying of water into S/C started.

Spraying of water into D/W started.

Spraying of water into S/C stopped.

Spraying of water into D/W stopped.

(Presumed) S/C vent closed S/C vent opened

(29)

 The trend in the dropping of the reactor pressure (Issues 6 and 9) was reproduced by

presuming that the HPCI had been operating continuously with flow rate adjustments made.

 It was not possible to eliminate the deviations of the PCV pressure values obtained from the analysis from the measured PCV pressure values during the periods in which the RCIC had been in operation, but the magnitudes of the deviations were reduced and the analysis

values became roughly the same as the measured values after commencement of the spraying of water into the S/C.

(The deviations may be attributable to the effects of the stratification of the S/C pool water, which cannot be analyzed with MAAP, and we think that further enhancement of

the analysis code should be made. )

 The reactor water level drop due to the HPCI shutdown led to the core damage, but RPV damage did not occur.

(According to our total evaluation based on the plant data obtained to date and other information, it is most likely that the RPV damage occurred. )

(30)

29 5. Conclusions

 We conducted an analysis based on the currently available information including

estimations (such as information on the operations performed by operators and estimations from the characteristics of the plant).

 As a result, we succeeded in reproducing the behavior of the plant at the time of the accident fairly accurately for the stages before the core melted.

 On the other hand, the RPV damage of Units 2 and 3 did not occur in this analysis, which is contrary to the observed facts (See the report of the “The Evaluation Status of Reactor

Core Damage at Fukushima Daiichi Nuclear Power Station Units 1 to 3 at Technical Work Shop.” (dated November 30, 2011)).

 The points made above shows the analysis capability limits of the current MAAP Code.

 Nuclear Emergency Response Headquarters Government-TEPCO Mid-and-long Term

Response Council has set up a sub-working team for investigating and analyzing the states of the inside of the reactors under the Research and Development Promotion Headquarters and started making efforts to further enhance the severe accident analysis codes including the MAAP Code.

 From now onwards, we will strive to further elucidate the development of the events and obtain a clearer picture of the states of the inside of the reactors by achieving higher

analysis accuracy levels utilizing the results of the abovementioned activities’ efforts and findings.

(31)

30 (Reference information)

Relationship between core state estimation and MAAP analyses

Core state estimation

(report dated November 30, 2011)

MAAP analyses

(Reports dated May 23 and December 22, 2011,

and this report)

Observed values of parameters including pressure

and temperature parameters

Analyses using other analysis

codes

Phenomena observed at work

sites

Investigations of the inside of the

cores using measuring equipment

Core state estimation involves total assessments of various kinds of information.

(The present MAAP analysis did not succeed in reducing the degree of uncertainty of the published core state estimation.)

原子炉圧力(MPa[abs]) Reactor pressure (MPa[abs]) Primary containment vessel pressure (MPa[abs])

MAAP Code-based Analysis

of the Development of the Events

at the Fukushima Daiichi Nuclear Power Station

March 12, 2012

Tokyo Electric Power Company

課題②

SRV main body

Steam

Steam

日時

原子炉水位（ m ）

日時

原子炉圧力(MP a[ ab s] )

日時

格納容器圧力( Mp a[ ab s] )

課題④ 課題⑤

日時

原子炉水位

日時

原子炉圧力

日時

格納容器圧力

課題⑥

課題⑦

日時

原子炉水位

日時 原子炉圧力

日時

格納容器圧力

Core state estimation

日時原子炉圧力