Incorporated administrative agency
Japan Nuclear Energy Safety Organization
Activities Related to Safety Regulations of
Spent Fuel Interim Storage at Japan Nuclear
Energy Safety Organization (JNES)
M.Kato, R.Minami and K.Maruoka
Japan Nuclear Energy Safety Organization (JNES)
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
1
Incorporated administrative agency
Japan Nuclear Energy Safety Organization
Contents
1. Current status of spent fuel interim storage in
Japan and Regulation Process
2. Research to investigate fundamental safety
functions of transport/storage cask for long term
storage
3. Research to investigate integrity of spent fuel
during storage
4. Safety Analysis
5. Ongoing and future activities
6. Summary
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
2
Incorporated administrative agency
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1. Current status of spent fuel interim
storage in Japan and regulation process
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
3
Incorporated administrative agency
Japan Nuclear Energy Safety Organization
Current Status of Spent Fuel Interim Storage in Japan
Project
Mutsu ISFSF (AFR)
max. 3,000 tU
Approval of license application : May 2010
Design and Construction Methods
Welding Inspection
Pre-Service Inspection
Source： HP of Recyclable-Fuel
Storage Company
Commencement of operation : July 2012
Chubu Electric Power
Hamaoka NPP :
max. 700 tU Metal Cask
Commencement of operation : 2016 FY
Source： HP of Chubu Electric Power
Kyushu Electric Power
ISFSF
Site investigation 2009 - 2011
Source： HP of Kyushu Electric Power
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
4
Incorporated administrative agency
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Flow of Nuclear Safety Regulation and Role of JNES(1/2)
Stage
NISA
Planning and
Design Stage
Safety Review
JNES
Technical Support :Data
for fundamental safety
function
Independent analysis to
validate safety assessment
by applicant
Construction
Stage
Approval of Design and
Construction Methods
Technical Support :
Preparation of technical
Criteria
Welding inspection
Preparation of inspection
procedure
Support
Order
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
5
Incorporated administrative agency
Japan Nuclear Energy Safety Organization
Flow of Nuclear Safety Regulation and Role of JNES(2/2)
Stage
NISA
Operation
Stage
Pre-Service Inspection
JNES
Preparation of Inspection
Procedure
Inspection (in part)
Approval of Operational
Safety Program
Operational Safety
Inspection
Annual Inspection
Continuous accumulation of
degradation phenomena
Preparation of Inspection
Procedure
Inspection (in part)
Confirmation of consignment
Support
Transportation method
confirmation
Order
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
6
Incorporated administrative agency
Japan Nuclear Energy Safety Organization
2. Research to investigate fundamental safety
functions of Transport/Storage Cask
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Scope of Research and Examination for Fundamental
Safety Function of Cask
Material property changes with time during long-term
storage and safety function
Material and Component
Safety Functions
◇Test for degradation of
◇Examination of
metal cask components
containment mechanisms
◇Test for degradation of
after long-term storage
concrete cask canister
・Drop Test(9m drop)
・Stress corrosion cracking of
・Thermal Test(fire
canister materials
condition)
(CRIEPI)
CRIEPI:Central Research Institute of Electric Power Industry
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Incorporated administrative agency
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Possible Degradation Phenomena of Metal Cask Component
Heat
Radiation
Atmosphere
NS(*1)
NS(*1)
Corrosion, SCC
(*2)
Overaging, Creep
NS(*1)
Corrosion, SCC
(*2)
Composition
change
Composition
change
-
relaxation
NS(*1)
Corrosion, SCC
(*2)
Cask body, Lid
(Carbon steel,
Stainless steal)
Basket
(Borated Aluminum
alloy)
Neutron shielding
(Resin, Propylene
glycol(PG)-water)
Seal boundary
(Metal gasket)
*1) NS: No Significance,
*2) Mainly due to degraded inner atmosphere
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
9
Incorporated administrative agency
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Test for degradation of metal cask components
Material Property(1/2)
Tests
Material of
Cask Body / Lid
(carbon steel,
stainless steel,
aluminum)
Material of
basket (borated
aluminum alloy)
Purpose
Confirmation of
corrosion
characteristic of
cask material due
to cask internal
atmosphere
deterioration
Confirmation of
long-term material
strength
characteristic of
basket material.
Main Results
In Iodine atmosphere assuming
1 % fuel failure , SCC did not
occur and corrosion is a little.
Mechanical, thermal properties etc.
were obtained when thermal ageing
or additional creep deformation
was applied. No important change
was observed.
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
10
Incorporated administrative agency
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Test for degradation of metal cask components
Material Property(2/2)
Tests
Purpose
Neutron shielding Confirmation of
materials (epoxy long term
resin, silicon resin, shielding
propylene glycol performance
water)
Metal gasket
( type: single or
double, material:
high nickel alloy
for spring,
aluminum for
outer jacket)
Confirmation of
relaxation change
due to thermal
aging
Main Results
Influence of radiation is negligible.
Degradation rate of both resins
caused by thermal ageing was
obtained.
An amount of relaxation due to
thermal aging was obtained.
Evaluation method of leak rate
from lid with relaxed metal gasket
were proposed, based on
experimentally obtained leak rate
trend data for displacement of lid.
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
11
Incorporated administrative agency
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Test for degradation of metal cask components
Safety Functions
Tests
Lid seal
performance
after 9m drop
Lid seal
performance
during fire
condition
(thermal test,
30 minutes 800
ºC )
Purpose
In drop accident in
transport after
long-term storage,
confirmation of
integrity of
confinement.
Main Results
Leak rate from lid was less than
1x10-5 Pa·m3/s. Evaluation method
for leak rate from lid with relaxed
metal gasket at drop event were
verified.
Applicability of DYNA-3D code to
estimate displacement of lid were
verified.
In fire accident in
transport after
long-term storage,
confirmation of
integrity of
confinement.
Maintaining containment safety of
lid with relaxed metal gasket
during fire event were confirmed.
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Test results
Material Property of Borated Al alloy for Basket
40
20
A3004
-H112
(1%B)
A6351
-T5
(1%B)
0
6N01
(5%B4C)
Annealing made strengths lower. Further, these
strengths were almost same if additional creep
deformation was provided.
Absorbing energy at impact test were almost same or
more than initial.
There was no important change for micro structure and
the other properties.
初期材
Initial
Annealed
過時効材
Annealed+Creepin
過時効+ｸﾘｰ
ﾌﾟ材
g
60
A5052
-H34
Test results
0.2% Proof Stress (MPa)
Subjects (Metals)
JIS H4080 A5052 H34 (No boron)
5wt%B4C Borated Aluminum Alloy (Base: JIS H4100 A6N01)
1wt% over Borated Aluminum Alloy (Base: ASTM A6351-T5)
1wt% Borated Aluminum Alloy (Base: ASTM A3004-H112)
Annealing Condition: (200 C, 250 C), (1,000hrs, 3,000hrs, 10,000hrs)
Testing Temperature:
Tensile Test (200 C, 250 C), Impact Test (-20 C),
Hardness (RT), Micro Structure
Modulus, Thermal Conductivity & Specific Heat,
Coefficient of Liner Expansion (RT, 100 C, 200 C, 250 C) 120
Mechanical Properties for Annealed and Creeping Metal
100
Annealing Condition: 250C, 1,000hours
Creep Deformation: about 0.1 % – about 1.0% (Max.)
80
Test Temperature (Tensile): 250 C
Comparison of proof strength (at 250 degree C)
Source： Interface issues between storage safety and post-storage
transport safety“Technical Meeting on Potential Interface Issues
in Spent Fuel Management”, 3–6 Nov 2009
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Incorporated administrative agency
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Test results
Neutron Shielding Materials
For Epoxy resin & Silicon resin;
irradiation tests of neutron or gamma
radiation, heating tests after irradiation,
heating tests etc.
Degradation condiution : 130 C to 170 C,
Max. heating time: 15,000 hrs.
4
Actual condition
estimated
Relations of weight loss and LMP (Larson･
Muller･Parameter)
LMP=T ( C + log t ) T: absolute
temperature of heating (K), C: constant, t:
heating time (hour)
Weight loss was estimated to occur by release
of oxide products of low molecular weight
from base materials and H2O due to
dehydrate reaction of tri-hydrate-alumina.
Heating was dominant for weight loss.
There was no synergistic effect of heating and
irradiation.
5
Loss (%)
Weight
重量減損(%)
Test results
130
C (Non-irradiated)
130℃非照射
150
C (Non-irradiated)
150℃非照射
170 C (Non-irradiated)
170℃非照射
130
C (irradiated)
130℃照射
150
C (irradiated)
170
C (irradiated)
150℃照射
170℃照射
95%信頼下限
95%信頼上限
6
3
2
1
0
15000
16000
17000
18000
劣化パラメータ
1.55*10-3 * LMP－25.3
19000
20000
21000
( C = 35 )
Degradation of Epoxy Resin
(in closed system with forced ventilation)
Source： Interface issues between storage safety and post-storage
transport safety“Technical Meeting on Potential Interface Issues
in Spent Fuel Management”, 3–6 Nov 2009
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
14
Incorporated administrative agency
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Results of 9m drop tests and thermal tests for lid
containment behavior and seal performance
Drop Tests using Full Size Cask
CASK Position
Horizontal Drop (1) & (2)
Vertical Drop with Lid Down
Corner Drop with Lid DownDrop
* For Horizontal (1), metal gaskets were prepared thermal degradation.
LMP=7400 was achieved.
Results
1.E-04
Horizontal Drop with Full size CASK
Leak rate of the secondary lid containment
system with relaxed metal gasket was estimated
lower than10-4 Pa･m3/sec on the drop of each
position.
Metal gasket elementary test results, radial
direction of dynamic, agreed to full size cask
drop.
Lid behavior in drop event was simulated well
by DYNA-3D code.
3/s)
３
Rate (Pa･m
Leak
漏えい率(Pa・m
/sec）
1.E-05
1.E-06
9G34(φ10、Ra=0.19、温度=7℃、速度=0～700mm/sec、B社)
9G33(φ10、Ra=0.15、温度=5℃、速度=0～700mm/sec、B社)
9G35(φ10、Ra=0.30、温度=11℃、速度=0～700mm/sec、B社)
9G36(φ10、Ra=0.15、温度=6℃、速度=0～700mm/sec、B社)
9G48(φ10、Ra=0.77、温度=7℃、速度=0～700mm/sec、B社)
9G47(φ10、Ra=0.65、温度=9℃、速度=0～700mm/sec、B社)
Results of “Degradation
9G43(φ10、Ra=0.79、温度=27℃、速度=0～700mm/sec、B社)
9G44(φ10、Ra=1.08、温度=30℃、速度=0～700mm/sec、B社)
tests for metal gasket“
9G45(φ10、Ra=2.90、温度=28℃、速度=0～700mm/sec、B社)
9G46(φ10、Ra=3.12、温度=29℃、速度=0～700mm/sec、B社)
9G41(φ10、Ra=0.22、温度=28℃、速度=0～500mm/sec、B社)
9G42(φ10、Ra=0.22、温度=30℃、速度=0～500mm/sec、B社)
B-2(φ10、Ra=0.30、温度=30℃、速度=0～700mm/sec、A社)
B-6(φ10、Ra=0.22、温度=30℃、速度=0～700mm/sec、A社)
Horizontal Drop (1)
水平落下1
水平落下2
Horizontal Drop (2)
上部垂直落下
Vertical Drop
上部コーナ落下
1.E-07
1.E-08
1.E-09
1.E-10
1.E-11
post-storage 0.0
Source： Interface issues between storage safety and
transport safety“Technical Meeting on Potential Interface Issues
in Spent Fuel Management”, 3–6 Nov 2009
1.0
Corner Drop
2.0
3.0
4.0
Radial Displacement
(mm)
横ずれ量(mm)
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
5.0
6.0
15
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3. Research to investigate integrity of
spent fuel during storage
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Background and JNES Test Plan for Evaluation of
Fuel Integrity
Technical Requirements in Japan
To prevent the failure of fuel
due to cladding thermal creep
To prevent the degradation of
cladding mechanical properties
Item
Survey and Planning
Creep Test
Creep Test
Technical Issues to be Evaluated
Thermal creep
» Hydride reorientation
» Irradiation hardening recovery
FY 2000 2001 2002 2003 2004 2005 2006 2007 2008
PWR48GWd/t, BWR50GWd/t
PWR55GWd/t, BWR55GWd/t
PWR48GWd/t, BWR50GWd/t
Creep Rupture Test
PWR48GWd/t, 55GWd/t
Hydride Effects Evaluation Test
BWR40GWd/t, 50GWd/t, 55GWd/t
» Hydride Reorientation Test
» Mechanical Property Test
PWR48GWd/t, BWR50GWd/t (330-420ºC)
（<330ºC）
Irradiation Hardening Recovery Test
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Spent Fuel Cladding Integrity Test - Summary
To develop the data for safety regulation, following mechanical property
tests were carried out from 2000 to 2008, using BWR and PWR fuel
cladding tubes irradiated in commercial power reactors in Japan.
(1) Thermal creep test, creep rupture test
» Threshold strain of transition to tertiary creep region is larger
than 1% for irradiated cladding.
» Creep equations were obtained for BWR and PWR claddings.
(2) Hydride reorientation and mechanical properties test
» Based on the experimental results, limit values of temperature
and stress in the dry storage were determined.
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Zry-4 cladding
strain
100
100
Threshhold strain to tertiary creep (%)
Threshold strain to tertiary creep (%)
Thermal Creep Test
eTh
tertiary
secondary
primary
10
10
Unirrad.
eTh : Threshold strain to tertiary creep
U nirradiated cladding
U nirradiated cladding 360℃
U nirradiated cladding 390℃
U nirradiated cladding420℃
U nirradiated cladding 420℃
Irradiated cladding 390℃
Irradiated cladding 420℃
Irradiated cladding 360℃
Iradiated cladding 390℃
Irradiated cladding 420℃
Irrad.
11
0.1
0.
1
10-7
1E-7
Irradiated cladding
: tertiary creep was not observed in the test
-6
10
1E-6
time
-5
10
1E-5
-4
10
1E-4
-3
10
1E-3
S econdary creep
(1/hr)
Secondary
creeprate
rate
(1/h)
The threshold strain of transition to tertiary creep was larger than
1% for irradiated cladding, 10 % for unirradiated cladding.
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Thermal Creep Test
Stress dependency of secondary creep rate
-4
1E-4
10
420ºC
Irradiated Zry-2 cladding
390ºC
Secondary creep rate (1/hr)
390℃
Creep equation
e  e t  e ps
420℃
10-5
1E-5
420℃
390℃
360℃
330℃
420℃-Calculated value
390℃-Calculated value
360℃-Calculated value
330℃-Calculated value
High
stress
region
High stress
region
nsH：7.7
360ºC
360℃
330ºC
-6
1E-6
10
330℃
Low stress
region
-7
1E-7
10
Low stress
region
nsL：1.3
e : Creep strain
e : Secondary creep rate
e sp: Saturated primary creep strain
t : Time
e  eL  eH
eL：Secondary creep rate
-8
1E-8
10
-9
1E-9
10
1E-4-4
10
(1)
BWR 50GWd/t type
-2
1E-3-3
101E-2
10
σ/E
s/E
(s: Hoop stress, E :Young’s modulus)
in the low stress region
eH：Secondary creep rate
in the high stress region
Creep rate was measured as parameters of stress and temperature using irradiated and
unirradiated fuel cladding tubes.
As results of creep test, it was shown that stress dependency of secondary creep rate was
different by stress regions, cladding types and irradiation.
Creep strain was expressed by equation(1) for BWR and PWR respectively.
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Hydride Effect Evaluation Test
Mechanical Property after Hydride Reorientation for PWR Zry-4
HRT340ºC 30ºC/h
HRT300ºC 30ºC/h
HRT275ºC 30ºC/h
HRT250ºC 30ºC/h
As-irradiated
Crosshead Displacement Ratio (%)
30
Crosshead displacement ratio :
index of ductility
25
20
15
As irradiated
10
5
48GWd/t type
HRT 300ºC, 115MPa, 30℃/h
0
0
50
100
150
200
HRT
Hoopreorientation
Stress (MPa) treatment (MPa)
Hoop stress during
hydride
Ring compression test was carried out to evaluate the effect of temperature
and stress on degradation of mechanical property.
Limit condition was determined by relative comparison with the value of
as-irradiated fuel cladding tube.
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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4. Safety Analysis
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Safety Analysis
Purpose：Though an independent analysis for the applicant analysis by
using analytical codes and/or methods for analyzing, to confirm whether
the applicant analysis results satisfy the criteria and whether the applicant
analysis is appropriate.
Input Date
・Open to the public data
・Offered data
Maintenance of analytical code
and method for analyzing
• Maintenance of mode of
analysis with high reliability
that reflects the latest finding
etc.
• Setting method of analytical
model and analysis condition
like how etc. to give method of
dividing analytical lattice and
boundary condition
• Verification analysis
Confirmed to satisfy the
criteria
Safety analysis
Confirmed that the
applicant analysis is
appropriate
• Check on applicant data
• Check on applicant analysis
condition
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Analytical Code for Independent analysis
Storage facility
Thermal
analysis
Fluid dynamics code
FLUENT
◆ Heat radiation analysis code
S-FOKS
◆
Cask
◆ Fluid dynamics code
FLUENT
Monte carlo code for neutron
Monte carlo code for neutron
and photon transportMCNP5
and photon transportMCNP5
Criticality  Monte carlo code for neutron  Monte carlo code for neutron
transport MVP-II
transport MVP-II
Analysis
 Japanese evaluated nuclear
 Japanese evaluated nuclear
data library JENDL-3.3
data library JENDL-3.3
 Impact and Structural
Structural
Analysis Code LS-DYNA
Analysis
Shielding
Analysis
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Incorporated administrative agency
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Temperature profile for postulated storage building
calculated by FLUENT coupling with SFOKS code
Importance of radiation heat transmission
■
Contributes to the except heat of the cask
Effect of decreasing cask surface
temperature at about maximum
20℃ compared with case only of
cooling by convection of air.
Heating of concrete
約17
■
Outlet
Intake
duct
The radiation from the barrel is
received, and the temperature rises
up to about the height 60℃.
約8.5
●
Metal cask
Concrete floor
約31
Temp. ℃
40
●
Heat radiation analysis code
■
S-FOKS code
Calculated by FLUENT coupling
with S-FOKS code
Metal cask
Concrete ceiling
35
29
Temperature profile calculated by FLUENT
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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5. Ongoing and future activities
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Ongoing and Future Activities
1.
Preparation of welding inspection procedure of canister (Corrosion
resistance stainless steel )
• Additional material properties were measured.
• Applicability of multi-layer PT and UT inspections for those
materials is under investigation.
2.
Preparation of technical criteria for design and construction method
approval
3.
Continuous improvement of safety analysis code and method
4.
Continuous accumulation of long term behavior of cask and spent fuel
• Demonstration test program for long term storage of PWR spent
fuel by utilities
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
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Summary
Activities related to safety regulations of spent fuel interim storage
at Japan Nuclear Energy Safety Organization is as follows.
Past:
• Fundamental safety function of metal cask during long term
storage.
• Seal performance under accident
• Integrity of spent fuel during long term storage
• Safety analysis code
Future:
• Support preparing criteria in regulations at the subsequent
stage
• Continuous improvement of safety analysis codes
• Continuous accumulation of long term behavior of cask and
spent fuel
International Conference on Management of Spent Fuel from Nuclear Power Reactors, May 31 – June 4, 2010
28