Demonstration Test Program for Long

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Demonstration Test Program for Longterm Dry
Storage of PWR Spent Fuel
IAEA-CN-178/08-03

• 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.

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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

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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.
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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
(55GWd/t fuel)
(48GWd/t fuel)
48GWd/t type fuel test 55 GWd/t type fuel
test Gas sampling
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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 1717 48GWd/t Fuel Assembly Type 1717 55GWd/t Fuel Assembly
Burn-up (MWd/t) 42,800 (past record) 55,000 (assumption)
Cooling period 19 years (as of October, 2012) gt10 years (as of October, 2022)
Cladding material Zircalloy4 MDA or ZIRLO
Remarks 15 empty fuel rods Non
Fuel rods used in PIE are never used for
long-term storage tests.
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2. Demonstration Test Program (3/5) Outline of
Test Container
Outline
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
Cross section
Mid-body
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.
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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
Start of Storage Test under Dry Condition
Inspection of fuel before storage test
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
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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.
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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 275C Around 230C (Gradually decrease with decrease in decay heat) Around 230C (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 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 During storage static position During earthquakes Acceleration of 1G
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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
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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 230C
  regarding to design value of actual casks.
• Gradual decrease of fuel temperature is simulated
  considering to the condition of actual casks.

Schematic drawing of Max. Temperature transition
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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) 55GWd/t (cooling for 10 years) .
Heat Load 547 W 1472 W (4551017 W)
Initial max. temperature of fuel cladding Approx. 250C (at 48GWd/t fuel assembly) Approx. 230C (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 230C or more.

Thermal Analyses of Test Container (during
loading of 4855GWd/t fuel assemblies)
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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 regulators standpoint. We will
  discuss its details in the future.

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Supplemental OHP Confirmation of containment
Schematic Drawing of Pressure Monitoring