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Challenge on

Severe Accident Simulation

Koji Okamoto

The University of Tokyo

[email protected]

日本原子力学会2015年秋の年会＠静岡September 9, 2015

「シビアアクシデント数値原子炉」

利用しやすいユーザーインタフェース

格納容器内外の

線量評価

炉内インベントリ（炉心燃焼計算）

解析結果の可視化処理機能

解析条件の設定や解析結果の

評価のための知識基盤

妥当性確認・精度評価（V&V)

再臨界

理論／物理

モデリング＆シミュレーション（Ｍ＆Ｓ）

小規模実験（Ｅ）

大規模複雑経済システム

大規模実験（Ｅ）

予測（外挿）

Ｖ＆ＶＶ＆ＶＶ＆ＶＶ＆Ｖ

不確かさ

高度化、３Ｄ化

高精度計測技術

Ass

essm

ent

V&Vにより、不確かさを低減し安全を確保

日本原子力学会シビアアクシデント評価研究会(2011-14)の目的

1. １Ｆのシビアアクシデントを評価検討し、SAMPSON

の高度化を通じて、１Ｆ内部の状況推定に資する

2. 今後のシビアアクシデント研究に対するアクションプラン

① SAMPSONコードによる1F推定の高度化コードを提供いただき、モデルの調査と評価を実施

②上記をサポートするV&Vデータ取得ロードマップ緊急性の高いデータを評価し、実験手法の提案

③ SA数値原子炉に向けたアクションプラン提案安全研究リストアップロードマップ化

5

fdada.info Current status of SA code for Fukushima

• Huge uncertainty on the simulation

– Uncertainty for Boundary conditions• Amount of water/sea-water supply, depressurization process,

heat removal by Tsunami flooding, etc.

– Uncertainty for Modeling• Fuel melting/solidification model

• Mass/Energy transfer between fuel and structure and so on.

– Uncertainty for Simulation Model• Mesh dependency, Multi-phase/dimensions, Radiation, etc.

6

Uncertainty for Debris Locations

Uncertainty for Source Term

Sample Simulation results by SAMPSON

High-pressure case Low-pressure case

7

Debris void fraction distribution in the core for Unit#1�High-pressure case:

keep 8MPa and cooling stops (same as Fukushima)

� Low-pressure case:

depressurize first and cooling stops (SAM strategy)

Absorber material in BWR

8

Stainless Steel

B4C powder

� BWR uses B4C as the absorber material

� B4C is filled in the stainless steel tube

� Stainless steel tubes are installed in

control blade

� BWR uses B4C as the absorber material

� B4C is filled in the stainless steel tube

� Stainless steel tubes are installed in

control blade

BWR fuel Control blade Absorber rod

http://www.toshiba.co.jp/nuclearenergy/english/business/reactor/abwr01_6.htm

http://www.world-nuclear.org/info/Nuclear-Fuel-Cycle/Conversion-Enrichment-and-Fabrication/Fuel-Fabrication/

Eutectic reaction (B4C/Stainless Steel)

9

Melting point

B4C ~2720K

Stainless Steel ~1730K

B4C/SS eutectic ~1500K

�Eutectic reaction of B4C and Stainless Steel starts at

around 1000K

�Melting temperature of eutectic material is lower than

the original materials

�When the temperature exceed 1500K, the eutectic

material starts to melt suddenly

Experimental Facility for tentative

experiment� B4C is filled between two Stainless Steel plates

� Heated by radiation heat

� Tungsten heaters are installed

� Ar atmosphere

10

Tun

gst

en

He

ate

r

Thermocouple

SS

B4C

po

wd

er

Front

TC

2mm 2mm 2mm 2mm

30

mm

1F1: core melt & relocation

11

Shroud temperature > 1250K

(1000C)

after 6 hours (over 1 hour)

• Buckling of rectangular columns under compression

– Buckling load decreases with increase of temperature.

– Lower buckling loads at high temp. than numerical predictions

Buckling failure under compression

Buckling load as a function of temperature

with numerical predictions`

Initial

column

Buckled

column

Captured images by

high speed cameraTest section

– Time to failure of the test columns was observed to be shorter.

– Several empirical correlations were used to predict failure time at various

temperature and load conditions.

• Larson-Miller Parameter

• Monkman-Grant relation

• Lateral deflection rate-failure time

Creep buckling at high temperatures

Experimental results and prediction

curves from Larson-Miller Parameter

0 m 60 m 66.4 m66.2 m

m: minute

Time evolution of a column deformation by creep buckling

169.1

min047.1 −⋅= δ&ft

Current status of SA code for Fukushima

• Huge uncertainty on the simulation

– Uncertainty for Boundary conditions• Amount of water/sea-water supply, depressurization process,

heat removal by Tsunami flooding, etc.

– Uncertainty for Modeling• Fuel melting/solidification model

• Mass/Energy transfer between fuel and structure and so on.

– Uncertainty for Simulation Model• Mesh dependency, Multi-phase/dimensions, Radiation, etc.

14

Uncertainty for Debris Locations

Uncertainty for Source Term

PIRT Project

• Objectives are– Develop Severe Accident Management Program

– Develop Probabilistic Risk Assessment Tools

– Understand the 1F accident precisely

– Estimate the debris distribution in 1F

• Phenomena Identification and its screening is most important process

• Ranking Table process depends on the target

• International Collaboration would be helpful for understanding the phenomena in comprehensive viewpoints

1F-PIRT• Objective① Estimation of the melted core status in the three

Fukushima-Daiichi NPPs

② Source term estimation improvement considering Fukushima-Daiichi NPP Accident.

• Phenomena Identification– Based on the knowledge of Severe Accident

Researches, the phenomena had been extracted.

– Discussion by experts, the key phenomena and/or the phenomena with large uncertainty had been selected.

• Organizer– Severe Accident Investigation Committee

the Atomic Energy Society of Japan(2011-14)

Fukushima-Daiichi PIRT papers

• N. Sakai, H. Horie, H. Yanagisawa, T. Fujii,

S. Mizokami & K. Okamoto,

“Validation of MAAP model enhancement for Fukushima

Dai-ichi accident analysis with Phenomena Identification

and Ranking Table (PIRT),”

J. Nucl. Sci. Technol., Vol.51, pp.951-963 (2014)

• S. Suehiro, J. Sugimoto, A. Hidaka, H. Okada,


“Development of the source term PIRT based on findings

during Fukushima Daiichi NPPs accident,”

Nuclear Engrg. and Design, Vol.286, pp 163-174 (2015)

Sakai et al., JNST (2014)

Suehiro et al., NED (2015)


Phenomena H M L K P U H & (P/U)

Core 179 73 75 30 60 108 10 51

Shroud Head 32 0 5 27 16 11 5 0

Standpipe & separator 32 4 4 24 15 9 8 3

Dryer 24 4 1 19 10 6 8 3

Upper Head 24 5 5 14 10 9 5 1

Main Steam Line 32 7 9 16 22 8 2 2

Upper Downcomer 31 3 12 16 17 9 5 0

Lower Dowmcomer 121 49 43 29 24 83 14 41

Lower Head 163 91 53 19 23 128 12 78

Recirculation Loop 36 2 4 30 13 16 7 2

Pedestal Cavity 140 69 35 36 21 100 19 63

Drywell 105 46 31 28 16 74 15 41

Drywell head 33 14 5 14 14 17 2 3

Vent to Wetwell 40 7 7 26 10 23 7 6

Wetwell 40 9 9 22 12 22 6 3

Isolaton Condenser 16 3 9 4 7 9 0 2

R/B Compartments 65 17 35 13 5 60 0 17

SGTS 2 0 2 0 1 1 0 0

Operation floor 30 4 11 15 0 30 0 4

Blowout panel 4 2 2 0 3 1 0 1

Spent Fuel Pool 6 0 2 4 0 6 0 0

Equipment Pool 1 0 0 1 0 1 0

TOTAL 1155 409 359 387 299 731 125 321


MAAP

Improvements


High Ranked Phenomena for Source Term

• Radionuclides release from molten fuel

• MCCI (Concrete erosion)

• Effects of B4C

– Chemical form

– Hydrogen generation

• Iodine release from R/B contaminated water

• Migration of radioactive material by the injection

water into the reactor

• Effects of seawater

– Chemical form


Fukushima-Daiichi PIRT papers

• N. Sakai, H. Horie, H. Yanagisawa, T. Fujii,


“Validation of MAAP model enhancement for Fukushima

Dai-ichi accident analysis with Phenomena Identification

and Ranking Table (PIRT),”

J. Nucl. Sci. Technol., Vol.51, pp.951-963 (2014)

• S. Suehiro, J. Sugimoto, A. Hidaka, H. Okada,


“Development of the source term PIRT based on findings

during Fukushima Daiichi NPPs accident,”

Nuclear Engrg. and Design, Vol.286, pp 163-174 (2015)



シミュレーションコードの役割

• 数値原子炉としてのＳＡコード

– アクシデントマネジメントの高度化

– リスク評価の不確かさ低減(レベル2PRA)

– 福島第一廃止措置へ不確かさ低減

• 基礎的事象のメカニズム評価

– 複雑現象のメカニズム解明

– 物理モデルの不確かさ低減理論／物理

モデリング＆シミュレーション（Ｍ＆Ｓ）

小規模実験（Ｅ）

大規模複雑経済システム

大規模実験（Ｅ）

予測（外挿）

Ｖ＆ＶＶ＆ＶＶ＆ＶＶ＆Ｖ

不確かさ

高度化、３Ｄ化

高精度計測技術

Ass

essm

ent

V&Vにより、不確かさを低減し安全を確保

SA Challenge

• 炉内溶融燃料挙動

–流路閉塞、共晶反応、酸化被膜、塩

• 構造材変形挙動

–流路閉塞、コリウム物性変化、崩落の有無

• 下部プレナムバウンダリ破損

–計測系配管、ＣＲＤＭ損傷、ドレイン配管

• 塩化ナトリウム

– コリウム物性への影響、ＭＣＣＩなどへの影響

SA Challenge

• 原子炉容器下部構造物との相互作用

– コリウムがCRDMなどに引っかかっている？

• ３次元MCCI挙動評価

– ドレンピット、ドレンライン、水冷却とガス

• ペデスタル内コリウム挙動

– コリウム流動、シェルアタック

• CFD応用

–格納容器内事象の総合的な評価

MCCI

• Sump Shape and Sump

drain

• Cooling characteristics

Depth

Width

Debris height

Invaded

MCCI

boundary

C/V steel wall

debris

Min. distance

1.02m

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