Document

Estimation of average burnup of damaged
fuels loaded in Fukushima Dai-Ichi Reactors
by using the 134Cs/137Cs ratio method
Nagoya University
Tomohiro ENDO, Shunsuke SATO,
Akio YAMAMOTO
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
1
Question

Why is radioactivity ratio, derived from
Fukushima Dai-ichi NPPs accident,
134Cs : 137Cs = 1 : 1 as of Mar. 11th, 2011?
0.253 [eV]
0.07
500 [keV]
0.06
14 [MeV]
0.05
0.04
0.03
0.02
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
Cs152
Cs151
Cs150
Cs149
Cs148
Cs147
Cs146
Cs145
Cs144
Cs143
Cs142
Cs141
Cs140
Cs139
Cs138m
Cs138
Cs137
Cs136m
Cs136
Cs135m
Cs135
Cs134
Cs133
Cs132
Cs131
Cs130
Cs129
0.00
Cs134m
0.01
Cs128
cumulative fission yield
0.08
2
Contents
 134Cs/137Cs
ratio method
 Numerical analysis
 Analysis of actually measured
134Cs/137Cs ratio


contaminated soils within the range of
100km from the 1F NPPs
Discussion
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
3
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 : 131I=130~160 [PBq], 137Cs=11~15 [PBq]
137Cs=4 [PBq]
Ocean
: 131I=11 [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
4
Purpose of this presentation

Estimation of burnup of damaged fuels
by using 134Cs/137Cs 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
5
134Cs/137Cs
ratio method
Radioactivity ratio of 134Cs/137Cs corresponds
to fuel burnup
 Convert measured 134Cs/137Cs ratio to burnup

1.8
Cs134/Cs137 [-]
measured ratio
1.2
0.6
estimated burnup
0.0
0
10000 20000 30000 40000 50000
burnup [MWd/t]
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
6
Production and depletion equation
for 134Cs and 137Cs
133Cs
134Cs
2 year
…
fission
137Cs
30 year
decay
capture
dN133 (t )
  c,133  N133 (t )   133  f 
dt
dN134 (t )
 134   c,134   N134 (t )   c,133  N133 (t )   134  f 
dt
dN137 (t )
 137   c,137   N137 (t )   137  f 
dt
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
7
Analytical solution 134Cs/137Cs ratio

134 N134 (t )   133   134   1   c,137  137  1  e  
 
 

 



N
(
t
)

1








1

e
N
(
t
)


1
137134
137 134 
137
 134 c,134 133 134  
 
137 N137(t) 2  


133

137
 c ,134  t
 1 ,
t 
 134
137  c ,137  t

1
t

c
,
133
 c ,133  t

134  c ,134  t ,
 e

1   c ,137  137
 

e


c
,
134

  

 1  137   137     1  e 137  c ,137  t
c ,134
c ,133
134  
 
 134   133 , 
134


   
 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
8
Modeling of numerical analysis

Detail information are classified due to
proprietary data
 U, Pu, and Gd enrichment/content splitting
in UO2 and MOX assemblies
 Fuel loading pattern
 Power, void, and temperature histories
 Simple model in the present research
 Pin cell geometry
 Assembly average values of 235U, Pu
 Typical core-averaged power, void, temp.
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
9
Loaded fuels and core averaged
specific power
fuel type
total number of
unit fuel asssemblies
1
400
2
548
3
548
UO2
8×8 STEP-II
UO2
9×9(A)
68
516
UO2
9×9(B)
MOX
8×8
thermal
power
[MW]
input
total heavy
specific
metal weight
power
[tHM]
[MW/tHM]
332
1380
70
20
548
2381
95
25
2381
97
25
32
[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
10
Dimensions & compositions of fuels
assembly
UO2
8×8
STEP-II
UO2
9×9(A)
UO2
9×9(B)
MOX
8×8
array
8×8
9×9
9×9
8×8
total number of fuel rods
60
66 (8)
72
60
assembly pitch [cm]
15.24
15.24
15.24
15.24
3.4
3.7
3.7
1.2
assembly-averaged 235U
enrichment [wt%]
assembly-averaged
Pu contents [wt%]
3.9
rod outer diameter [cm]
1.23
11.2
1.1
1.23
thickness of cladding [cm]
0.086
0.71
0.07
0.086
effective fuel length [cm]
371
371(216)
371
355
cell pitch [cm]
1.63
1.44
1.44
1.63
fuel pellet diameter [cm]
1.04
9.6
0.94
1.04
gap between pellet
and cladding [cm]
0.02
0.2
0.02
0.02
density of fuel pellet [%TD]
97
97
97
95
fuel rod
material cladding
only assembly
average values
are published
fuel rod information
is sufficient to carry
out pin-cell calc.
zircalloy-2 zircalloy-2 zircalloy-2 zircalloy-2
[6] http://www.nsc.go.jp/shinsashishin/pdf/1/ho007.pdf
[7] http://www.pref.fukushima.jp/nuclear/info/pdf_files/100714-2.pdf
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
11
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
12
Comparison of 134Cs/137Cs 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
13
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
14
Numerical results of 134Cs/137Cs ratio
among calculation codes
99(B) fuel, VF=40%, 25 [MW/tHM]
Cs [-]
1.0
0.02
134
137
1.3
Cs/
1.5
radioactivity ratio of

0.8
0.5
SRAC2006/PIJ (ENDF-B/VII.0)
0.3
MVP-BURN (ENDF-B/VII.0)
SCALE6.0/TRITON (ENDF-B/VII.0)
0.0
0
5
10
15
20
25
30
burnup [GWd/tHM]
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
15
Void fraction and nuclear data library
effects for 134Cs/137Cs 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

axial node number

24
23
22
21
20
19
18
17
16
15
14
13
12
11
10
9
8
7
6
5
4
3
2
1
0%
20%
40%
60%
80%
VF
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
16
Numerical results of 134Cs/137Cs ratio
for different VF & nuclear data library

99(B) fuel, 25 [MW/tHM]
Lib. Dif.
2.0
1.5
radioactivity ratio of
134
Cs/
137
Cs [-]
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)
0.05
VF. Dif.
0.3
1.0
0.5
0.0
0
5
10
15
20
25
30
burnup [GWd/tHM]
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
17
Comparison of 134Cs/137Cs ratio
among fuel type

Depends on specific power
 Difference between UO2 and MOX
2.25
1.50
1.25
134
137
Cs [-]
←others
Cs/
1.75
radioactivity ratio of
←UO2, 25 [MW/tHM]
2.00
UO2(8x8), 20MW/tHM
1.00
UO2(9x9B), 20MW/tHM
0.75
UO2(9x9B), 25MW/tHM
0.50
UO2(9x9A), 25MW/tHM
0.25
MOX(8x8), 25MW/tHM
0.00
0
10
20
30
40
50
burnup [GWd/tHM]
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
18
Estimation formula for fuel burnup
burnup [GWd/tHM]
30

SRAC2006/PIJ
with JENDL-4.0
 Weighting 134Cs &
137Cs by tHM for
each cores
25
20
15
VF= 0%
VF=40%
VF=70%
10
5
0
0.0
0.3
0.6
0.9
1.2
1.5
radioactivity ratio of 134Cs/137Cs [-]
1.8
18.04 x  0.8321 x 2  3.162 x 3 (VF  0%)

B( x)  15.94 x  1.409 x 2  2.779 x 3 (VF  40%)
14.36 x  1.785 x 2  2.676 x 3 (VF  70%)

Nov. 17th, 2011, 2011 Symposium on Nuclear Data
19
Contamination densities of 134Cs &137Cs

Contaminated soils within the range of 100 km
from the Fukushima Dai-ichi NPPs
134Cs
137Cs
÷
[8] http://www.mext.go.jp/b_menu/shingi/chousa/gijyutu/017/shiryo/__icsFiles/afieldfile/2011/09/02/1310688_2.pdf
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
20
250
Frequency distribution of 134Cs/137Cs

150
0
1.5
1.4
1.3
1.2
1.1
1.0
0.9
0.8
0.7
0.6
0.5
50
100
frequency
200
0.996±0.07
0.6
0.8
1.0
1.2
1.4
1.6
Cs134/Cs137 as of 2011/3/11
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
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
21
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
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
22
Discussion

Core burnup by plant process computer
1
cycle
3.8
burnup
[GWd/tHM] 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 coreaveraged value
 Possible causes


Postulated core meltdown process
Once-burned fuel
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
23
lower
upper
Postulated core meltdown process
center

Damages of Fuel
Assemblies (FAs)
progressed from center to
peripheral region

FAs loaded in peripheral
region are typically 4th and/or 5th-burned fuel

Averaged burnup of
damaged fuel would be
lower than that of core
averaged value
[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
24
Infinite neutron multiplication factor
Once-burned fuel

~10 [GWd/tHM]
for 1 cycle(1 year)
once-burned FAs,
which have relatively
high power density
due to burnout of
burnable poison,
may be highly
damaged due to
higher decay heat.
Burnup [GWd/tHM]
Nov. 17th,
2011, 2011 Symposium on Nuclear Data
[12] CASMO-4/SIMULATE-3
コードシステムによるBWR実機炉心解析に関する報告書,
JNES/SAE05-029
25
Conclusion

In the present research, estimated burnup is
17.2±1.5 [GWd/tHM] by using 134Cs/137Cs ratio
method for contaminated soils

VF effect in depletion calculation has a major
impact on 134Cs/137Cs 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.
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
26
Thank you for your attention
We sincerely thank all of
researchers that are involved in
the measurement and analysis of
radiation dose and radioactivecontamination map project
supported by MEXT
Nov. 17th, 2011, 2011 Symposium on Nuclear Data
27

Download Report