1. Demonstration Test Program for Long–term Dry Storage of PWR Spent Fuel
IAEA-CN-178/08-03
Demonstration Test Program for Long–term Dry Storage of PWR Spent Fuel
2 June 2010
M. Yamamoto,
The Japan Atomic Power Company
The Kansai Electric Power Co., Inc.
Kyusyu Electric Power Co., Inc.
Mitsubishi Heavy Industries, Ltd.

2. Contents 1. Introduction 2. Demonstration Test Program
Test Overview and Process
Fuel Assemblies for Test
Outline of Test Container
Verification Method of Fuel Integrity
Confirmation during Storage Tests
3. Designing of Test Container
Current Knowledge and Experience
Simulated Environment of Actual Casks
4. Summary

3. 1. Introduction
Mutsu interim spent fuel storage facility in Japan is preparing for the maximum 50-year storage of spent fuel in dry metal casks for both transportation and storage.
To reduce risk of radiation exposure to workers and waste materials, the facility has no hot cells, and the spent fuel will be confirmed for their integrity indirectly by monitoring casks during storage and transported after the storage without opening the cask lid.
Lots of fuel cladding integrity investigations in Japan
Lots of demonstrations &
experiences in overseas
Dry storage experiences
of BWR fuel in Japan
Long-term storage test for “fuel integrity” in domestic research facility
to accumulate knowledge and experience on long-term integrity
of PWR spent fuel during dry storage.
To make assurance doubly sure on safety of transportation after storage.

4. 2. Demonstration Test Program (1/5) Test Overview and Process
Time Schedule of Demonstration Test of PWR Fuel Storage
Fiscal year
2009
2010
2011
2012
2013
–2022
2023
–2032
2033
–2042
2043
–2052
- - -
Planning
& Designing
Manufacture
& Preparation
Storage test & Inspection
Planning
Designing
Safety analysis
Licensing
Manufacturing of test container
Thermal test
Preparation & Fuel inspection
Loading to container
(48GWd/t fuel)
(55GWd/t fuel)
48GWd/t type fuel test
55 GWd/t type fuel test
Gas sampling

5. 2. Demonstration Test Program (2/5) Fuel Assemblies for Test
Up to two spent fuel assemblies (Type 48GWd/t and 55GWd/t ) will be stored.
48GWd/t : Some of the fuel rods were used for PIE tests, and now it is stored in the pool of the hot laboratory in Tokaimura (NDC).
55GWd/t : a proper spent fuel will be prepared in the future.
Fuel assemblies Assumed for Tests
Type 17×17 48GWd/t Fuel Assembly
Type 17×17 55GWd/t Fuel Assembly
Burn-up (MWd/t)
42,800 (past record)
≤55,000 (assumption)
Cooling period
19 years
(as of October, 2012)
>10 years
(as of October, 2022)
Cladding material
Zircalloy–4
MDA or ZIRLO
Remarks
15 empty fuel rods*
Non
*Fuel rods used in PIE are never used for long-term storage tests.

6. 2. Demonstration Test Program (3/5) Outline of Test Container
item
Description
Compo-nents
- Lid
(Steel, Resin,
Double metal gasket)
- Body
(Steel, insulator, Resin)
- Basket
(Steel, Boron-Al)
- Outer thermal insulator
Size
- Height :
Approx. 5.2m
- Outer diameter
Approx. 2.2m
Contents
Max. 2 PWR spent fuel assemblies
Cover gas
Helium (negative pressure)
Lid
Outer thermal insulator*
Inner thermal insulator
Mid-body
Cross section
Inner container
PWR spent fuel assemblies
Basket spacer
(Boron-Al)
Neutron shield
Basket
(Stainless steel)
Trunnion
*Note: Outer thermal insulator installed at loading only 48GWd/t F/A is removed when 55GWd/t fuel assembly is added.

7. Inspection of fuel after storage test Investigation of cause
2. Demonstration Test Program (4/5) Verification Method of Fuel Integrity
Loading to test container*
* The following inspections of sampled cover gas are to be carried out at the start of storage test after fuel loading;
– Kr-85 radioactivity analysis
– Gas composition analysis
[ 48GWd/t fuel assembly]
Inspection of fuel before storage test
Start of Storage Test
under Dry Condition
– Visual inspection of fuel assembly
10 years
Analysis and monitoring during storage test
Loading to test container*
[ 55GWd/t fuel assembly]
Inspection of fuel before storage test
– Kr-85 radioactivity analysis
– Gas composition analysis
– Monitoring of surface temperature of test container
– Monitoring of containment boundary pressure of test container
– Visual inspection of fuel assembly
Increase of Kr-85 level
End of Test
Inspection of fuel after storage test
Suspension of Test
Investigation of cause
Flow Diagram of Test Program
– Visual inspection of fuel assembly

8. 2. Demonstration Test Program (5/5) Confirmation during Storage Tests
Sampling of cover gas in test container
- Confirm detection of fuel leakage
- Induction of cover gas into sampling pod
- Scheduling every 5 years
- Radioactive gas (Kr-85) analysis with a Ge detector
- Gas components analysis with a mass spectrometer
Temperature monitoring
- Estimate temperature history of fuel rods
- Installation of thermocouples on the outer surface in the middle area.
- Calculation of the fuel rods temperature with a previously-verified assessment tool by thermal performance tests.
Pressure monitoring
- Confirm maintenance of containment of the test container
- Monitoring of helium gas pressure at the lid boundary.
- Installation of pressure gauges to a buffer tank leading to gap of double metal gaskets.

9. Evaluation of Degradation Events
3. Designing of Test Container (1/4) Current Knowledge and Experience
Evaluation of Degradation Events
Conditions to be considered
Technical Evidence
Actual Conditions of
Stored Cask
Test Conditions (target)
Thermal degradation
No embitterment due to hydride
reorientation, failure due to creep
strain, recovery of irradiation
hardening, or stress corrosion crack under 100MPa or less
circumferential stress at 275°C
Around 230°C
(Gradually decrease with decrease in decay heat)
(Gradually decrease with decrease in
decay heat)
Chemical degradation
Negligible oxidation/hydrogen
absorption during storage (inert gas atmosphere) compared to that
during in-core irradiation
He gas atmosphere
Moisture:
10% or less
Radiation degradation
Negligible neutron irradiation
influence during storage
Saturation of mechanical strength
due to neutron irradiation at
relatively low burn-up (around 5GWd/t)
Burn-up of stored fuel:
Maximum 47GWd/t
Burn-up of contained
fuel:
5GWd/t or more
Mechanical degradation
Maintenance of integrity under
normal test conditions of transport (free drop) (Acceleration :20 to 45G)
During storage:
static position
During earthquakes:
Acceleration of 1G

10. Radiation degradation Mechanical degradation
3. Designing of Test Container (2/4) Simulated Environment of Actual Casks
Chemical degradation
The test container is filled with helium gas having negative pressure as with actual dry cask cavity.
Vacuum drying operation is carried out before backfilling of helium gas.
Amount of moisture is confirmed.
Mechanical strength of cladding tubes shows saturation and ductility shows slow deterioration at low burn-up (around 5GWd/t).
Test fuel Burn-up is 42.8GWd/t. (Irradiation dose is 1021 to 1022n/cm2)
The test container is statically positioned in a vertical direction.
Radiation degradation
Mechanical degradation

11. Schematic drawing of Max. Temperature transition
3. Designing of Test Container (3/4) Simulated Environment of Actual Casks --- Temperature
Thermal degradation
The maximum temperature of fuel cladding tubes during the storage test is set as around 230°C regarding to design value of actual casks.
Gradual decrease of fuel temperature is simulated considering to the condition of actual casks.
Test Time （year）
Maximum temperature (°C)
230
210
55GWd/t fuel assembly
48GWd/t fuel assembly 3 2 1
Schematic drawing of Max. Temperature transition

12. 3. Designing of Test Container (4/4) Simulated Environment of Actual Casks --- Temperature
Heat load and max. temperature of cladding at initial test conditions
Beginning of test
Addition of 55GWd/t fuel assembly
Loaded fuel assemblies
48GWd/t
(cooling for 19 years)
48GWd/t (cooling for 29 years) & GWd/t (cooling for 10 years) .
Heat Load
547 W
1472 W ( W)
Initial max. temperature of fuel cladding
Approx. 250°C
(at 48GWd/t fuel assembly)
Approx. 230°C
(at 55GWd/t fuel assembly)
Thermal analyses were conducted for estimation of max. temperature of fuel claddings covered with He gas.
The obtained temperatures will meet the aimed temperature around 230°C or more.
Thermal Analyses of Test Container (during loading of 48&55GWd/t fuel assemblies)

13. 4. Summary
Some Japanese utilities are planning to conduct a long-term storage test for up to 60 years by placing PWR fuel assemblies in a test container simulating temperature and internal gas of actual casks to accumulate knowledge and experience on long-term integrity of PWR spent fuel assemblies during dry storage.
The storage test plans such as test methods and inspection items, and container design have been prepared. In the future, safety analyses, licensing and manufacturing of the test container are to be done, and the storage test of 48GWd/t fuel assembly will start in fiscal 2012.
Thermal design of the test container is important. Its temperature is controlled with thermal insulators and heat-transfer performance is confirmed by heat transfer tests at the completion of the container.
Others Japan Nuclear Energy Safety Organization (JNES) plans to participate in this test from a regulator’s standpoint. We will discuss its details in the future.

14. Supplemental OHP Confirmation of containment
Inert gas inlet / outlet
Lid part
Web data logger
Wireless
Server
Non–controlled area
Controlled area V3 V2 V1 V4 V5
V10 V6 V7 V8 V9 PC
Buffer tank P1 P2 T1 T2
Schematic Drawing of Pressure Monitoring