Transcript handout
Estimation of average burnup of damaged fuels loaded in Fukushima Dai-Ichi Reactors by using the 134 Cs/ 137 Cs ratio method
Nagoya University
Tomohiro ENDO, Shunsuke SATO, Akio YAMAMOTO Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Question
Why is radioactivity ratio, derived from Fukushima Dai-ichi NPPs accident, 134 Cs : 137 Cs = 1 : 1 as of Mar. 11th, 2011?
0.08
0.07
0.06
0.05
0.04
0.03
0.02
0.01
0.00
0.253 [eV] 500 [keV] 14 [MeV] Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Contents
134 Cs/ 137 Cs ratio method Numerical analysis Analysis of actually measured 134 Cs/ 137 Cs ratio contaminated soils within the range of 100km from the 1F NPPs Discussion Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Overview of 1F NPPs accident
Station blackout accompanied with loss of cooling capability and loss of ultimate heat sink due to excessive tsunami (~15m) caused by M9.0 earthquake at 14:46 Mar. 11th, 2011 Severe core damage in units 1-3, confinement capabilities (RPV, CV) are partially damaged Release of radioactive nuclides to environment Atmosphere : 131 I=130~160 [PBq], 137 Cs=11~15 [PBq] Ocean : 131 I=11 [PBq], 137 Cs=4 [PBq] [1] 原子力安全・保安院 , ( 平成 23 年 6 月 6 日 ) [2] 原安全委員会 , 第 64 回原子力安全委員会資料第 3 号 ( 平成 23 年 8 月 24 日 ) [3] H. Kawamura, et al., JNST,
48
[11], p.1349
–1356 (2011) Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Purpose of this presentation
Estimation of burnup of damaged fuels by using 134 Cs/ 137 Cs ratio method Effectiveness of estimated burnup Health effects due to released radioactive nuclides from 1F NPPs Isotopic composition depends on fuel burnup Especially important for unsurveyed isotopes Burnup credit for criticality safety for discharging process of fuel-debris Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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134
Cs/
137
Cs ratio method
Radioactivity ratio of 134 Cs/ 137 Cs corresponds to fuel burnup Convert measured 134 Cs/ 137 Cs ratio to burnup 1.8
measured ratio 1.2
0.6
estimated burnup 0.0
0 10000 20000 30000 40000 50000
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Production and depletion equation for
134
Cs and
137
Cs
fission 133 Cs 134 Cs … 137 Cs decay 2 year 30 year capture
dN
133 (
t
) c , 133
N
133 (
t
) 133 f
dt dN
134 (
t
)
dt
134 c , 134
N
134 (
t
) c , 133
N
133 (
t
) 134 f
dN
137 (
t
)
dt
137 c , 137
N
137 (
t
) 137 f
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Nov. 17th, 2011, 2011 Symposium on Nuclear Data
Analytical solution
134
Cs/
137
Cs ratio
134
N
134
N
137
N
( (
t t
) ) 134
N
137 ( (
t
) 133
t
) 133 137 137 134 1 2 1 1 1 134 1 137 c , 134 c , 137 , c c , 134 137 137 133 134 c , 133 137 137 134 1 1 c , e e e 1 3 4 1 3 7 c 1
t
, 1 3 3 e
t
c c , , 1 3 4 1 3 7
t
t
t
1 134 ,
t
e 1 3 4 134 c 1 , 1 3 4 c , c , 134
t
133
t
, , Rigorously, microscopic reaction rates and fission yield depend on fuel burnup Numerical solution can be solved by Bateman’s method Matrix exponential method Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Modeling of numerical analysis
Detail information are classified due to proprietary data U, Pu, and Gd enrichment/content splitting in UO 2 and MOX assemblies Fuel loading pattern Power, void, and temperature histories Simple model in the present research Pin cell geometry Assembly average values of 235 U, Pu Typical core-averaged power, void, temp.
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Loaded fuels and core averaged specific power
input unit total number of fuel asssemblies UO2 8×8 STEP-II 1 2 3 400 548 548 68 fuel type UO2 9×9(A) UO2 9×9(B) 332 548 516 MOX 8×8 32 thermal power [MW] total heavy metal weight [tHM] specific power [MW/tHM] 1380 2381 2381 70 95 97 20 25 25 [4] TEPCO homepage, http://aoisora.org/genpatu/2011/tepco_data/20110409151130/atomfuel01-j.html
[5] TEPCO homepage, http://www.tepco.co.jp/nu/f1-np/intro/outline/outline-j.html
MOX is ~10-years storage fuels and loaded in this cycle for the first time Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Dimensions & compositions of fuels array total number of fuel rods UO2 8×8 STEP-II 8×8 60 UO2 9×9(A) 9×9 66 (8) UO2 9×9(B) 9×9 72 MOX 8×8 8×8 60 assembly assembly pitch [cm] assembly-averaged 235 U enrichment [wt%] assembly-averaged Pu contents [wt%] rod outer diameter [cm] thickness of cladding [cm] 15.24
3.4
1.23
0.086
371 15.24
3.7
11.2
0.71
371(216) 15.24
3.7
1.1
0.07
371 15.24
1.2
3.9
1.23
0.086
355 effective fuel length [cm] fuel rod cell pitch [cm] fuel pellet diameter [cm] gap between pellet and cladding [cm] density of fuel pellet [%TD] material cladding 1.63
1.04
0.02
1.44
9.6
0.2
1.44
0.94
0.02
1.63
1.04
0.02
97 97 97 95 zircalloy-2 zircalloy-2 zircalloy-2 zircalloy-2 only assembly average values are published fuel rod information is sufficient to carry out pin-cell calc.
[6] http://www.nsc.go.jp/shinsashishin/pdf/1/ho007.pdf
[7] http://www.pref.fukushima.jp/nuclear/info/pdf_files/100714-2.pdf
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Other input conditions
Void fraction (VF) of coolant : 40% Temperature Fuel ： 900 [K] Cladding: 600 [K] Moderator ： 560 [K] These values are typical BWR core parameters Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Comparison of 134 Cs/ 137 Cs ratio among various calculation codes Deterministic code SRAC2006/PIJ (Collision probability method) SCALE6.0/TRITON (Discrete ordinate method) Monte Carlo code MVP-BURN total number of histories: 5000 × 80 for each burnup step With same nuclear data library: ENDF-B/VII.0
SRAC2006/PIJ : 107 energy groups SCALE6.0/TRITON : 238 energy groups MVP-BURN : continuous energy Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Calculation scheme of SCALE6.0/TRIRON Deterministic code for neutron transport and depletion calculations Resonance Calculation 2-D discrete ordinate method for neutron transport calculation Fuel depletion and decay calculation Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Numerical results of 134 Cs/ 137 Cs ratio among calculation codes 9 9(B) fuel, VF=40%, 25 [MW/tHM] 1.5
0.02
1.3
1.0
0.8
0.5
0.3
SRAC2006/PIJ (ENDF-B/VII.0) MVP-BURN (ENDF-B/VII.0) SCALE6.0/TRITON (ENDF-B/VII.0) 0.0
0 5 10 15 burnup [GWd/tHM] 20 Nov. 17th, 2011, 2011 Symposium on Nuclear Data 25 30
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Void fraction and nuclear data library effects for 134 Cs/ 137 Cs ratio SRAC2006/PIJ Void Fraction (VF) 0% for lower part 40% for average 70% for upper part Nuclear data library JENDL-4.0
ENDF-B/VII.0
24 23 22 21 20 19 18 17 16 8 7 6 5 4 3 2 1 15 14 13 12 11 10 9 0% 20% 40% VF 60% Nov. 17th, 2011, 2011 Symposium on Nuclear Data 80%
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Numerical results of 134 Cs/ 137 Cs ratio for different VF & nuclear data library 9 9(B) fuel, 25 [MW/tHM] 2.0
1.5
VF=0% (JENDL-4.0) VF=40% (JENDL-4.0) VF=70% (JENDL-4.0) VF=0% (ENDF-B/VII.0) VF=40% (ENDF-B/VII.0) VF=70% (ENDF-B/VII.0) 1.0
Lib. Dif.
0.05
VF. Dif.
0.3
0.5
0.0
0 5 10 15 burnup [GWd/tHM] 20 25 Nov. 17th, 2011, 2011 Symposium on Nuclear Data 30
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Comparison of 134 Cs/ 137 Cs ratio among fuel type Depends on specific power Difference between UO 2 2.25
and MOX 2.00
1.75
←UO 2 , 25 [MW/tHM] ←others 1.50
1.25
1.00
0.75
0.50
0.25
0.00
UO2(8x8), 20MW/tHM UO2(9x9B), 20MW/tHM UO2(9x9B), 25MW/tHM UO2(9x9A), 25MW/tHM MOX(8x8), 25MW/tHM 0 10 20 30 burnup [GWd/tHM] 40 50 Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Estimation formula for fuel burnup
30 25 20 SRAC2006/PIJ with JENDL-4.0
15 10 5 VF= 0% VF=40% VF=70% Weighting 134 Cs & 137 Cs by tHM for each cores 0 0.0
0.3
0.6
0.9
1.2
1.5
radioactivity ratio of 134 Cs/ 137 Cs [-] 1.8
B
(
x
) 18 .
04 15 14 .
94 .
36
x x x
0 .
1 8321 .
409
x
1 .
785
x
2 2
x
2 3 .
162 2 .
779
x
3 2 .
676
x
3
x
3 (VF 0%) (VF 4 0%) (VF 7 0%) Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Contamination densities of 134 Cs & 137 Cs Contaminated soils within the range of 100 km from the Fukushima Dai-ichi NPPs 134 Cs 137 Cs ÷ [8] http://www.mext.go.jp/b_menu/shingi/chousa/gijyutu/017/shiryo/__icsFiles/afieldfile/2011/09/02/1310688_2.pdf
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Frequency distribution of 134 Cs/ 137 Cs 0.996
± 0.07 1.5
1.4
1.3
1.2
1.1
1.0
0.9
0.8
0.7
0.6
0.5
0.6
1.6
0.8
1.0
1.2
Cs134/Cs137 as of 2011/3/11 1.4
Estimated burnup : 17.2
± 1.5 [GWd/tHM] [9] http://www.mext.go.jp/b_menu/shingi/chousa/gijyutu/017/shiryo/__icsFiles/afieldfile/2011/09/02/1310688_1.pdf
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Plant data at 14:00 Mar. 11th (1F#3) Alarm recording data includes numerical summaries of BWR plant process computer burnup data [10] http://www.tepco.co.jp/nu/fukushima-np/plant-data/f1_3_Keihou3.pdf
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Discussion
Core burnup by plant process computer burnup [GWd/tHM] 1 cycle 3.8
core-averaged 25.8
unit 2 2* 23.2
3 4.2
21.8
*Hard to read due to low quality of published pdf file Estimated burnup(17.2
± 1.5 [GWd/tHM]) is nearly equal to but slightly lower than core averaged value Possible causes Postulated core meltdown process Once-burned fuel Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Postulated core meltdown process
Damages of Fuel Assemblies (FAs) progressed from center to peripheral region FAs loaded in peripheral region are typically 4 th and/or 5 th -burned fuel Averaged burnup of damaged fuel would be lower than that of core averaged value center [11] 東京電力株式会社福島第一原子力発電所の事故に係る 1 号機、 2 号機、 3 号機の炉心の状態に関する評価 報告書 , JNES-RE-2011-0002 Nov. 17th, 2011, 2011 Symposium on Nuclear Data
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Once-burned fuel
once-burned FAs, which have relatively high power density due to burnout of burnable poison, may be highly damaged due to higher decay heat.
~10 [GWd/tHM] for 1 cycle(1 year) Burnup [GWd/tHM] Nov. 17th, 2011, 2011 Symposium on Nuclear Data , JNES/SAE05-029
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Conclusion
In the present research, estimated burnup is 17.2
± 1.5
[GWd/tHM] by using 134 Cs/ 137 Cs ratio method for contaminated soils VF effect in depletion calculation has a major impact on 134 Cs/ 137 Cs ratio More precise evaluation requires more detail information about fuel assemblies’ data loaded in 1F NPPs: histories and distributions of the specific power and the void fraction are strongly desired.
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Thank you for your attention
We sincerely thank all of researchers that are involved in the measurement and analysis of radiation dose and radioactive contamination map project supported by MEXT
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