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Dry storage systems and aging management

H.Issard,

AREVA TN, France

IAEA TM 47934

LESSONS LEARNED IN SPENT FUEL MANAGEMENT

Vienna, 8-10 July 2014

AREVA TN

TN International

IAEA TM 47934 Lessons Learned in Spent Fuel Management Vienna, 8-10 July 2014 2

Summary

Dry storage systems and AREVA Experience

Aging management: Deployed systems / inspection

mitigation

New aging management program in the long term

Aging analysis for confirmation of safety of extended

interim storage

Conclusion

TN International


AREVA Experience dry storage systems

Experience of used fuel storage designed by AREVA: Metal cask

Metal casks (TN 24):

� Dual purpose : storage and transport

� Forged carbon steel shell

� Bolted lids & metal seals

� Long Industrial Experience :first cask in 1985

� Monitoring: temperature, interlid pressure

� Inspections and Surveillance records : no

significant event

� Safety review for storage extension.

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Experience of used fuel storage designed by AREVA : canister system

»Canister based systems (Nuhoms®):

�Storage System components:

� Dry shielded canister

� Horizontal storage module

� Transfer cask

�Auxiliary equipment

�Long Industrial Experience : first system in 1989

�Inspections and Surveillance records : no significant event

�visual inspection

�monitoring: temperature

�Safety review for extension: lead canister inspection.

2014: AREVA has received the license for Nuhoms® MP 197HB transportation

cask, for the transport of high burnup fuels up to 62 GWd/MTU

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Experience of used fuel storage designed by AREVA : vaults

Vaults (Cascad type):

�Vaults: storage only

�storage pits. UNF are placed in leak-tight

welded canisters (single element or multi

element).

�Long Industrial Experience : in operation since

1990

�Storage conditions monitored

�Safety records : no significant event

�Safety review for extension

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►Leaking fuels

►Capsule Canister for Handling and Storage of Fuel Rod Capsules

►Dimension for handling and storage similar to a Fuel Assembly

Spacer plate

Bottom end piece

Transport

head

Fuel rod

capsules

Corner lining strip

Tie rod Transport head

Spacer plate

Corner lining strip

Tie rod

Bottom end piece

For use at a

PWR:

loading capacity

up to 108 Fuel

Rod Capsules

For use at a BWR:

loading capacity up to

33 Fuel Rod Capsules

• AREVA´s technology in

encapsulation of

defective Fuel Rods.

• Drying method : meet

the strict requirements of

maximum water content

for storage.

• Can guarantee a

reduction of

residual water

content << 1 g per

fuel rod.

TN International


Operational Experience dry storage systems

Nuclear operators and cask vendors: broad experience in the world in used fuel storage

Aging management program and deployed systems / inspection mitigation

� Periodical inspections

• visual for containment corrosion

• Surveillance of air inlets and outlets

• dose rate

� Permanent monitoring

• Storage parameter measurements : temperature, inter-lid pressure

• gamma and neutron dose rate

� Inspections of content after cask opening

• Seal inspection, Moisture, Fuel integrity

� Time limited aging analysis : ex. creep or corrosion tests

� Analysis based on surrogates : ex. He build up

Lessons learned :

� No accident and no event,

� But very limited experience of cask opening

� Some aging issues need further justification (influence of drying)

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Dry storage feedback from inspections after cask opening

US – Post Irradiation Examination (PIE – in hot cell) of spent fuel

assemblies stored 15 years in Castor-V/21 cask

� no significant degradation to cask or FA’s

UK – PIE of PWR-UO2 fuel (58 GW d/tU) stored 20 years under air

� no change in the cladding oxide thickness

Japan – Post irradiation examination

� 2 MOX BWR rods irradiated to 20 GW d/tU were stored for 20 years in an air

capsule

• no release of fission gas/He and no change in fuel/cladding microstructure

� Integrity inspections at Tokai (with metal cask) and Fukushima-Daiichi (with TN24

cask) performed after 5 and 10 years of storage - 52 BWR assemblies (≈30GW

d/tU)

• No release of inner gas (Kr-85) and no defect observed on the spent fuel assemblies

(visual checking)

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Development of new aging management program

For years, a storage period of up to 40 years was considered as ‘long-term’ and sufficient in considering decisions and deployment of back-end fuel cycle and/or final waste management options.

Extension of dry storage : Some consider an extension of the storage duration significantly, even beyond one century.

Address the need to store used fuel with higher burn ups (62 Gwd/t for PWR or 70 Gwd/t for BWR ) and MOX fuel

Used fuel should be retrievable for further use

=> Safety of interim storage in the long term ; need of new aging management program

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Aging management to assure long term performance

Focus on ensuring primary safety functions

� Criticality control

� Confinement/containment

� Shielding

Focus on components providing safety functions

� Canister/cask and transport package= Primary for confinement/containment =

Can provide criticality control

� Concrete storage module = shielding

� Canister/cask internals and fuel = defense in depth for containment criticality

control

Focus on activities to provide assurance

� Monitoring

� Inspection

� Mitigation

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Aging management program to ensure long term performance

Aging management program (AMP) updated with results of new analysis of degradation of dry storage components:

� Cask or canister material

� concrete

� sealing system

� coatings

� neutron shielding materials

� neutron poison materials

Actions

� Tests, models

� Inspection program

� Mitigation

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Aging analysis for confirmation of safety for extended storage

Analysis of cask or canister material

� Main function : containment of used fuel.

� Submitted to ambient atmosphere and humidity.

� Most important degradation process : atmospheric Stress Corrosion Cracking

(SCC), or Chlorine induced SCC.

� Inspections performed on existing canisters have shown no signs of SCC.

� However, investigations are underway for Chlorine induced SCC evaluation for

welded canister system in marine environment: amount of Chlorine in ambient

atmosphere, assessment of deliquescence of salt, evaluation of corrosion

through inspection of existing storage systems, modelling corrosion and SCC.

� Mitigation techniques (stress reduction), coatings or material changes could

be considered for increasing weld durability.

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Inspection to support Analysis

� Inspection of cask or canister material for salt deposition

� Analysis Chlorine induced SCC

Tool carrier in use during training

SaltSmartTM device (Louisville solution, USA)SaltSmartTM delivry tool

in dry run on canister surface

TN International



Analysis of concrete

� Main function: radiological shielding and physical protection for the

Canister against a wide range of postulated natural hazards.

� The NUHOMS® system provide an independent, passive system with

substantial structural capacity to ensure the safe dry storage of used fuel

assemblies.

� The long term concrete degradation is associated with temperatures and

radiation levels: chemical degradation, carbonation, corrosion of

embedded steel, coupled mechanisms, dry-out and thermal degradation of

mechanical properties. Results from nuclear and non nuclear industrial

experience and inspection are satisfactory.

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Analysis of sealing system

� For metal casks, the function of the metal seals is the

containment

� Long term issues are the corrosion of bolts and the

corrosion of metal seals.

� The behaviour of these components has been studied

in storage conditions with satisfactory results by CEA

(France) and CRIEPI (Japan). These studies cover long

term resistance (leak tightness tests) of metal seals

with Al or Ag outer jackets. Corrosion tests on such

seals have also shown good resistance .

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Analysis of neutron shielding materials

� Function : to protect the public and the operators form the neutrons radiations

� Industrial experience : no increase of the dose rate during storage.

� In-service neutron shielding ageing: irradiation or thermal oxidation processes.

� Accelerated aging tests: various temperatures and O2 partial pressure

� To predict the long term properties, a non-empirical model is applied, taking

into account the diffusion-limited oxidation.

• The model simulates very confidently weight losses (which are then converted into

hydrogen atoms loss) and oxidation profiles.

• good agreement between experimental results and simulated data.

0

5

10

15

20

25

30

35

90 100 110 120 130 140 150 160 170

Temperature (°C)

Hy

dro

ge

n lo

ss

(%

)

1 year

3 years

7 years

10 years

20 years

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Analysis of neutron poison materials

� Main function : to prevent criticality.

� Evaluation of creep : Industrial experience (casks in vertical position) has

shown no observation of creep of neutron absorbers due to control of

temperatures over extended storage times.

� Wet corrosion and blistering (for Boral, a porous neutron poison material)

• water entering pores in the material during loading

• Vaporization during vacuum drying

• Blistering may cause dimensional changes affecting criticality considerations

� Material used now : no corrosion/blistering

� Thermal aging effects: decrease in tensile and yield strength.

� Embrittlement due to radiation exposure: unlikely

� Nowadays neutron poison materials degradation negligible

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Used fuel behaviour in dry storage: many IAEA technical

documents

Creep of Used fuel cladding (Zr Alloys)

� At dry storage temperatures between 300 and 400 °C, the cladding

undergoes strain due to creep.

� Temperature decreases continually during dry storage, no significant creep

strain expected

� Creep strain is largely determined by fuel rod internal pressure and fuel rod

temperature time history.

� Provided that the maximum cladding temperature does not exceed 400°C,

creep under storage will not cause gross rupture of the cladding.

� Creep self limiting phenomenon: not a critical threat to used fuel integrity

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Used fuel behaviour in dry storage: many

IAEA technical documents

Hydrogen effects – embrittlement

� Storage temperature will decrease to the point

that hydrides precipitate in the Zr cladding .

� Hydrogen uptake depends on material,

irradiation history and oxide thickness.

� Hydrogen uptake and hydride precipitation

decrease the ductility of irradiated zirconium

alloys.

� The precipitation of radial hydrides can reduce

significantly the ductility.CEA –R-6084

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Pressure increase due to He-generation

� Used fuel pellets: production of helium from alpha decay during storage

� For very long storage periods (hundreds of years), the production of

helium in MOX fuel becomes comparable to the amount of fission gases

produced during reactor irradiation.

� For the considered storage duration (several decades), the pressure

increase is limited

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New AMP developments and studies

Aging/degradation Stressor and risk Solution mitigation Need further work

UNF Creep T, P, Bu : rupture Temperature limit No

UNF oxydation T, BU, gas composition :

rupture in transport

Inert gas No

UNF H2-effects Clad alloy, T, P, Bu:

Rupture in transport

Drying temperature limit

Time limit for transport

Complementary tests

UNF DBTT Clad alloy, T, P, Bu :

rupture in transport

Drying temperature limit

Time limit for transport

Complementary tests

Canister stainless steel Chlorides, resid stress:

SCC rupture

Control, Prevent salt

deposit, SCC resistant

welds

New coating, selection

treatment

Cask closure lid Chlorides, resid stress:

SCC rupture

Control, Prevent salt

deposit, SCC resistant

New coating, selection

treatment

Concrete Corrosion Temperature limit

Material selection

No

Aluminium from basket Creep Temperature limit

Material selection

Complementary tests

seals Corrosion Assessment of corrosion No

Coatings corrosion Assessment of corrosion Complementary tests

Neutron shielding Loss of properties Temperature limit Complementary tests

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Conclusion

The overall experience is a safe and reliable storage performance.

Aging management plan associated with extension of storage

New investigations and tests underway for better understanding and managing the degradation of casks or canister materials

Used fuel behaviour is a key issue for the management of the back end of the nuclear fuel cycle.

Confirmatory demonstration is needed (already started) with high burn up fuel.

Involvement of AREVA in these actions

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This document and all information contained herein is

intellectual proprietary to AREVA TN / TN International.

They shall not be disclosed, in whole or in part, without its

authorisation.

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