1

　 Overview of Fukushima-Accident Analysis

ERMSAR 2012, Cologne (Germany)

March 21 – 23, 2012

JNES

Masanori FUKASAWA

2

Contents

1. 1F 1※ Accident Analyses (Plant behavior) at JNES

2. Plant Behavior Analysis using MELCORA) Results at IAEA Ministerial Conference (June 2012) and Problems

B) Revised Analysis

3. Primary System Behavior during IC operation

4. Hydrogen Mixing and Explosion in Reactor Building (R/B)

5. Conclusions

※1F: Fukushima Daiichi

3

Published Analyses and Evaluations ”Report of Japanese Government to the IAEA Ministerial Conference on

Nuclear Safety,” June 2011. (June Report), JNES-RE-2011-0002.Documents of Hearings at Nuclear Safety Commission (NSC) and Nuclear

and Industrial Safety Agency (NISA).Analyses and Evaluations submitted to NISA (published on JNES web).

Accident Analysis Plant Behavior FP Release•Getting chronology information together

•Event tree analysis of the accident

•Possibility of recriticality•Reactivity constraint by sea water

•Time before fuel damage in SFP

•Salt precipitation

•MELCOR analysis•Primary system behavior during IC operation

•Hydrogen mixing and explosion

•MCCI in case water injection stops

•Possibility of PCV failure by H2 deflagration

•H2, O2 concentration

•FP release and dose evaluation

• Influence in case water injection stops

•FP release and EPZ •Estimation of FP release and dose based on monitoring data

•FP release in case of venting

1. 1F Accident Analyses (Plant behavior) at JNES

4

Analytical configuration:

Code: MELCOR1.8.5

2.　 Plant Behavior Analysis using MELCOR

6 volumes of primary system4 volumes of containment5 volumes of reactor building

to simulate FP transfer.Junctions of S/R valves,

vacuum breaker, PCV leak, W/W vent

Further (not depicted), activated cooling systems and assumed leak to simulate transports of steam, coolant and FP. 1F1： IC1F2： RCIC1F3： RCIC, HPCI

Objectives: To figure out plant behaviors of 1F1 – 3 and enhance safety measures.

Primary boundarySecondary boundary

Pedestal

蒸気ドーム

下部プレナム

Upper plenum

Core

Bypass

Do

wncom

er

RPV

FHB

4F

3F

2F

1F

Environment

Blowout panelPCV

Vent pipe

W/W

RPV failure

W/W

vent

ADS

S/RV

Vacuu

m

breaker

R/B

Leakag

e

SGTS

D/W

Steam dome

Lower plenum

Primary boundarySecondary boundary

Pedestal

蒸気ドーム

下部プレナム

Upper plenum

Core

Bypass

Do

wncom

er

RPV

FHB

4F

3F

2F

1F

Environment

Blowout panelPCV

Vent pipe

W/W

RPV failure

W/W

vent

ADS

S/RV

Vacuu

m

breaker

R/B

Leakag

e

SGTS

D/W

Steam dome

Lower plenum

5

①IC stop、② water injection、③W/W vent open、④W/W vent close、⑤ Sea water injection、⑥Increase of PCV leakage

1F1: Lower coolant injection caseWater injection (fire pump, F/P) by 3/15 is 88 m3

2.A) 　 Result of June Report and Problems (1F1)

IC actuation is limited and water level decreases at an early stage.Core melts before alternate water injection.RPV failure is calculated at 5 hrs.Most core is calculated to melt and slump to PCV floor.　RPV failure timing (MELCOR default model calculates early failure.)Actuation of W/W vent at 1st attempt (3/12 10:17).

Analytical results

Problems

0.0

1.0

2.0

3.0

4.0

5.0

6.0

7.0

8.0

-4000

-2000

0

2000

4000

6000

0 12 24 36 48 60 72 84 96

(M

Pa)

炉圧

(m

m)

水位

(h)経過時間

(○,△ )実測値1F1

炉圧

TAF

RPV水位

3/ 12 3/ 13 3/14

実時刻

① ② ③

④

⑥⑤

W/W vent

Date

Wat

er le

vel (

mm

)

P/S

pre

ssur

e (M

Pa)

Time (hr)

○,△：Measured dataP/S Pressure

Water level

P/S pressure and water level

6

1F2: In case PCV confinement maintains

D/W pressure could not be reproduced in case PCV confinement maintained.→PCV leakage was assumed.

①RCIC manual actuation②SBO

③ Change of RCIC water source from CST to S/P

④RCIC stop⑤Sea water injection⑥S/R valve1 open⑦S/R valve open⑧explosion

D/W pressure increases due to temperature rise of S/P.

Water source of RCIC is switched from CST to S/P.

Assumption of analysis in June Report

2.A) 　 Result of June Report and Problems (1F2)

D/W pressure

0.0

0.2

0.4

0.6

0.8

1.0

0 12 24 36 48 60 72 84 96

M

Pa

圧力

（）

(h)経過時間

D/ W (○)圧力実測値

D/ W圧力( )破損無し

2Pd

1Pd

実時刻

①②

③ ④ ⑥⑤

⑧⑦

3/ 120:00

3/ 130:00

3/ 140:00

3/ 150:00

Date

D/W

pre

ssur

e (M

Pa)

Time (hr)

○：Measured data

7

2.A) 　 Result of June Report and Problems (1F2)

1F2: Lower coolant injection case with PCV leakage Water injection (F/P) by 3/15 is 213 m3

D/W pressure is well simulated on assumption of PCV leakage (50 cm2) at an early stage.

RPV failure is calculated at 80 hrs because water injection by F/P is not enough.

Higher FP release due to assumed early PCV leakage

Analytical results

0.0

0.2

0.4

0.6

0.8

1.0

0 12 24 36 48 60 72 84 96

M

Pa

圧力

（）

(h)経過時間

D/ W (○)圧力実測値2Pd

1Pd

実時刻

①②

③ ④ ⑥⑤

⑧⑦

D/ W圧力

3/ 120:00

3/ 130:00

3/ 140:00

3/ 150:00

D/W pressure

①RCIC manual actuation, SBO, ② ③ Change of RCIC water source from CST to S/P,

④RCIC stop, Sea water injection, S/R valve1 open, ⑤ ⑥⑦S/R valve open, explosion⑧

Date

D/W

pre

ssur

e (M

Pa)

Time (hr)

○：Measured data

D/W Pressure

8

2.A) 　 Result of June Report and Problems (1F2)

0.0

0.2

0.4

0.6

0.8

1.0

0 12 24 36 48 60 72 84 96

M

Pa

圧力

（）

(h)経過時間

D/ W (○)圧力実測値2Pd

1Pd

実時刻

①②

③ ④ ⑥⑤

⑧⑦

D/ W圧力

3/ 120:00

3/ 130:00

3/ 140:00

3/ 150:00

D/W pressure

①RCIC manual actuation, SBO, ② ③ Change of RCIC water source from CST to S/P,

④RCIC stop, Sea water injection, S/R valve1 open, ⑤ ⑥⑦S/R valve open, explosion⑧

Date

D/W

pre

ssur

e (M

Pa)

Time (hr)

○：Measured data

Measured D/W pressure does not increase even S/R valve opened. On the other hand, pressure increases in calculation.

D/W pressure at this stage is not simulated due to assumed PCV leakage.

Some heat removal (instead of PCV leakage) possibly occurred by then.

Problems

1F2: Lower coolant injection case with PCV leakage Water injection (F/P) by 3/15 is 213 m3

D/W Pressure

9

1F3: Lower coolant injection caseWater injection (F/P) by 3/17 is 369 m3

Much H2 is produced due to water level decrease when S/R valve opens.

RPV failure is calculated at 79 hrs because sea water injection is not enough.

FP is released to environment through W/W vent.

Analytical results

①RCIC actuation、② RCIC stop, HPCI actuation, HPCI ③ ④stop, S/R valve open, W/W vent open, water injection, ⑤⑥W/W vent close, W/W vent open, Sea water injection, ⑦ ⑧⑨W/W vent close, water injection、⑩～⑭W/W vent open⇔close

2.A) 　 Result of June Report and Problems (1F3)

0.0

1.0

2.0

3.0

4.0

5.0

6.0

7.0

8.0

-4000

-2000

0

2000

4000

6000

0 24 48 72 96 120 144

(M

Pa)

炉圧

(m

m)

水位

(h)経過時間

(○ △ )、 実測値

1F3

炉圧

TAF

RPV水位

実時刻L-2 L-8と の間で推移

① ②③

⑤ ⑨～

⑩ ⑭⑪ ⑫ ⑬④

3/ 12 3/13 3/14 3/15 3/16 3/17 3/18

S/R Valve open

P/S pressure and water level

Date

Wat

er le

vel (

mm

)

P/S

pre

ssur

e (M

Pa)

Time (hr)

○,△：Measured data

P/S Pressure

Water level

10

During RCIC operation, D/W pressure is underestimated. (inverse trend to 1F2)

Measured pressure drops when HPCI actuates.

During HPCI operation, water level is not clear.

Amount of produced H2

Explosions of 1F3 and 1F4 are possibly attributed to H2 produced in 1F3

2.A) 　 Result of June Report and Problems (1F3)

D/W pressure

Problems

0.0

0.2

0.4

0.6

0.8

1.0

0 12 24 36 48

M

Pa

圧力

（）

(h)経過時間

D/ W (○)圧力実測値

D/ W圧力

2Pd

1Pd

実時刻

① ②③

④ ⑥ ⑨～

3/ 120:00

3/ 130:00

3/ 1212:00

3/ 1312:00

⑤

Date

D/W

pre

ssur

e (M

Pa)

Time (hr)

○：Measured data

D/W Pressure

11

RCIC steam exhaust pipe

Hot water

Hot water flow

Major problems; D/W pressure underestimation during RCIC operation and pressure drops after HCPI actuation.

Latest information and examination RCIC continuous operation using

return line to CST.→Assume S/P thermal stratification by

RCIC exhaust steam. (see Figs.)• Simulated by upper/lower S/P nodes and RCIC exhaust to the upper.

• HPCI initiated, steam exhausted to lower node assuming lower temp. of the water near HPCI exhaust pipe.

PCV spray during HPCI Similar pressure transition between

P/S and S/C after 42.4 hrs. →Assume RPV failure at this time.

2.B) Revised Analysis (1F3: Analytical Conditions)

12

Improved matching with measured data

D/W pressure increases during RCIC operation

Depressurization due to S/C spray （ No depressurization if thermal stratification not considered because of lower spray flow rate)

Remained problemsModeling of S/P thermal

stratification; investigation using CFD

Further investigation is needed for PCV leakage and W/W vent, which have large influence on FP release.

0

2

4

6

8

0.0

0.2

0.4

0.6

0.8

1.0

0 10 20 30 40 50 60

(MPa

)原

子炉

圧力

PCV

(MPa

)圧

力

(h)経過時間

原子炉圧力

D/ W圧力→

実時刻3/ 120:00

3/ 130:00

3/ 140:00

2.B) Revised Analysis (1F3: Analytical Result)

Date

D/W

Pre

ssur

e (M

Pa)

P/S

pre

ssur

e (M

Pa)

Time (hr)

P/S Pressure

D/W pressure

○：Measured data

P/S and D/W pressures

S/C Spray

13

2.B) Revised Analysis (1F2: Analytical Conditions)

Major problems; D/W pressure after S/R valve open. (Some heat removal instead of PCV leakage)

Latest information and examination Tsunami water flooded at a depth of boots length in RCIC room

(same level as S/P torus room) at 1:00, 3/12 and increased at 2:00.*

→Assume S/P heat removal by flooding water• 60% heat of RCIC exhaust steam is removed.

Early PCV leakage is not assumed. Instead;• Small leakage at 70 hrs because measured D/W pressure

slightly decreases. • Enlargement of leakage at 90 hrs when large pressure drop

is measured. RCIC injection rate is adjusted to simulate time when water

level comes down to TAF. Assume S/P thermal stratification (Same as 1F3).

*TEPCO, “Report regarding factual information related to the investigation results of the accident situation at Fukushima Daiichi Nuclear Power Plant,” Dec. 22, 2011.

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1.0

2.0

3.0

4.0

5.0

6.0

7.0

8.0

9.0

-4000

-2000

0

2000

4000

6000

0 20 40 60 80 100

(MPa

)炉

圧

(mm

)コ

ラプ

スト水

位

(h)経過時間

(○,△ )実測値

炉圧

TAF

ダウンカマ水位

実時刻3/ 150:00

3/ 120:00

3/ 130:00

3/ 140:00

BAF

E[2]

G[5]

F[3]

C

A,B

D[1]

H[10]

P/S pressure and water level

A:RCIC manual actuation, B:SBO, C:Change of RCIC water source from CST to S/P, D[1]:RCIC stop, F[3]Sea water injection, E[2]:S/R valve1 open, G[5]:S/R valve open, H[10]explosion

2.B) Revised Analysis (1F2: Analytical Result)

P/S pressure is also simulated by adjusting RCIC injection rate

Date

Wat

er le

vel (

mm

)

P/S

pre

ssur

e (M

Pa)

Time (hr)

P/S Pressure

Water level

○△：Measured data

15

0.0

0.2

0.4

0.6

0.8

1.0

0.0

0.2

0.4

0.6

0.8

1.0

0 20 40 60 80 100

(MPa

)炉

圧

D/W

(MPa

)圧

力 (h)経過時間

(○,△ )実測値

炉圧

D/ W圧力

実時刻3/ 140:00

3/ 130:00

3/ 120:00

3/ 150:00

0.6 cm2 of leakage

Enlargement of leakage (32 cm2)

2.B)　 Revised Analysis (1F2: Analytical Result)

P/S and D/W pressures

High D/W pressure is reproduced.

Calculated pressure increase becomes lower and consistent with measured data.Steam through S/R

valve flows to lower level of S/P, whose temperature is lower due to thermal stratification.

Date

Wat

er le

vel (

mm

)

D/W

pre

ssur

e (M

Pa)

Time (hr)

P/S Pressure

D/W pressure

○△：Measured data

16

3. Primary System Behavior during IC operation

0.0

1.0

2.0

3.0

4.0

5.0

6.0

7.0

8.0

3/ 11 14:45 3/ 11 15:00 3/ 11 15:15 3/ 11 15:30 3/ 11 15:45 3/ 11 16:00 3/ 11 16:15

DAY

Rx P

ress

(MPa

,g）

RELAP5RecorderChart

EarthquakeSCRAM

IC start

IC stop

IC manualoperation

SR/ V

Isolation Condenser (IC) is a unique system for reactor cooling in unit-1, and worked at the initial stage of the accident. (ceased due to AC/DC valve power loss by the tsunami)

RELAP5/mod3 analyses were performed to investigate the IC behavior. IC functioned properly to the original design. (Not impaired by the earthquake) Sensitivity analysis shows that the core uncovery could have been avoided by

continued operation of IC after the tsunami.

IC system (Unit-1)Initial stage of the accident (after earthquake)

Assuming IC continued operation

Reactor pressure

-6000

-4000

-2000

0

2000

4000

6000

8000

3/1114:30

3/1115:00

3/1115:30

3/1116:00

3/1116:30

3/1117:00

3/1117:30

3/1118:00

3/1118:30

3/1119:00

3/1119:30

DAY

Abo

ve T

AF (m

m）

EarthquakeSCRAM

Assuming IC continuedoperation after tsunami Feed seconary

coolant

Water level above TAF

17

Hydrogen gas concentration

4. Hydrogen Mixing and Explosion in R/B

JNES is conducting analyses of hydrogen gas mixing and detonation in Reactor Buildings (R/Bs) for investing explosion phenomena during Fukushima accident.

MELCOR for hydrogen source evaluation , FLUENT(CFD code) for hydrogen gas transport and mixing, and AUTDYN for structural analysis of detonation

The objectives are to better understand the phenomena that took place in Unit 1 and Unit 3, and assess and improve the methods and tools.

CFD model of Reactor Building Mixture gas velocity

18

Some results from the analysis of detonation at unit 3

4. Hydrogen Mixing and Explosion in R/B

Debris Velocity R/B Pressure

The amount of hydrogen gas leaked into R/B is estimated to be approximately 1 ton.

If it is assumed that the leakage took place at S/C or D/W’s penetration, overall detonation behavior is well reproduced.

With initial velocity 70 m/s debris is supposed to reach at about 250 from the top of R/B at 7.1 seconds.

This photo is quoted from Fukushima-chuo TV This photo is quoted from TEPCO web

19

• JNES has been conducting various analyses of the Fukushima accident.

• Plant behavior analysis using MELCOR improved by assuming S/P thermal stratification and latest information for 1F2 and 1F3.

• P/S behavior analysis of 1F1 using RELAP5/mod3 shows IC functioned properly to the original design. (Not impaired by the earthquake)

• Detonation analysis with the assumption that leakage took place at S/C or D/W’s penetration well estimates overall R/B behavior of 1F3.

5. Conclusions