Management of Spent Nuclear Fuel

Stefano Caruso National Cooperative for the Disposal of Radioactive Waste (NAGRA)

Hardstrasse 73, 5430 Wettingen, Switzerland

“A sneak peek of the International Conference on the Management of Spent Fuel from Power Reactors” Webinar, 3rd May 2019

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Management of Spent Nuclear Fuel

S. Caruso

The safe, secure, reliable and economic management of spent fuel arising from nuclear power reactors iskey for the sustainable utilization of nuclear energy

Swiss concept: SNF flow to final disposal

Spent Fuel poolReactor Dry Interim Storage Cask Transportation

Nagra Surface Facility SF unloading SF loading Canister welding Final emplacement in geological repository

“A sneak peek of the International Conference on the Management of Spent Fuel from Power Reactors” Webinar, 3rd May 2019

3

Management of Spent Nuclear Fuel

S. Caruso

Main aspects related to storage, transportation, encapsulation and disposal of the spent fuel and the highlevel waste need to be previously addressed to develop the: Operational and long-term safety concepts Optimised design of the encapsulation facility and repository layout

Ageing management is also needed to deal with a relative long-term storage time

RD&D programs are needed to consolidate the scientific basis for supporting safe assessment and designrequirements. E.g. two categories:

I. Spent Fuel Characterisation and fuel behaviour at the time of emplacement (pre-disposal)

• Dose rate, decay heat, fuel assembly integrity during handling/encapsulation in disposal canisters

II. Spent fuel Evolution on the long-term (post-disposal)• Radionuclides inventory for large set of nuclides• Criticality Safety Assessment - with Burnup Credit application and scenarios evolution

“A sneak peek of the International Conference on the Management of Spent Fuel from Power Reactors” Webinar, 3rd May 2019

Relevant phenomena affecting fuel cladding performance after storage

Creep

Hydrides build up

Delayed Hydride Cracking

Air Oxidation

Stress Corrosion Cracking

Hydrogen pick up

Reduction of priority for dry storage (D

oE)

High priority for dry storage

Potential for fuel rod failures

S. Caruso4

Influencing factors: Burnup, Temperature, Pressure

“A sneak peek of the International Conference on the Management of Spent Fuel from Power Reactors” Webinar, 3rd May 2019

I. Nagra / JRC Karlsruhe: Experimental program on SNF integrity

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3-point bending test

Impact test

Force vs. displacement

Impact sequence

Post-test examinations:1. H2 measurement2. Metallography3. Fractography4. Fuel mass release5. Particle size distribution

0 5 10 15 20 25 30 35 40

0.5

0.6

0.7

0.8

0.9

1.0

Kinetic Energy Loss

Divided by Max of Distance (pixel) Divided by Max of Distance (mm)

Frame

Nor

mal

ized

Dist

ance

(pix

el/p

ixel

)

0.5

0.6

0.7

0.8

0.9

1.0

Nor

mal

ized

Dist

ance

(mm

/mm

)

Prin

cipa

l Axi

sHigh-speed imaging

Beam

Numerical model

Pellet/Cladding

𝜎𝜎𝑓𝑓𝑓𝑓 = 𝑓𝑓(𝐵𝐵𝐵𝐵, ε)

1. Effective mechanical properties

2. Constitutive laws3. Failure model4. Stress analysis5. 𝜎𝜎 = 𝑓𝑓(𝜎𝜎𝑓𝑓𝑓𝑓)

S. Caruso “A sneak peek of the International Conference on the Management of Spent Fuel from Power Reactors” Webinar, 3rd May 2019

I. Mechanical tests – 3point bending & Impact

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Data acquisition (sensors): - Load, deflection, pressure

Low BU sample retains ductility

High BU sample → higher stiffness and strength- PCI might indicate difference in flexural moduli

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.00.00

0.17

0.33

0.50

0.67

0.83

1.00

Load

(a.u

.)

Displacement (a.u.)

Low BU High BU

Impact recorded by a high speed camera- Spallation (ejection) of the outer oxide layer- Crack initiation position, propagation and morphology- Fuel fragment release progression

S. Caruso

Bending tests Impact tests

“A sneak peek of the International Conference on the Management of Spent Fuel from Power Reactors” Webinar, 3rd May 2019

II. Determination of radionuclide inventory in Fuel Matrix and Cladding

Task: Determining inventory with a fuel depletion integrated to anactivation model

How: Development of SCALE /TRITON MC-libraries for fuel andcladding material

Validation: e.g. KIT measurements of C-14 in Zircaloy-4 claddingand stainless steel plenum spring sampled from a fuel rod segmentirradiated in the Swiss Gösgen PWR

Scope: Estimation of radionuclide inventory

3860 mm

3860 mm

S. Caruso7

Zry-4 14C [Bq/g]

Measured (KIT) 3.7(±0.4)×104

Calculated(Nagra) 3.6×104

C/E 0.97

Build-up of C-14: 1) N-14+n -› C-14 + p; 2) C-13+n -› C-14 + γ; 3) O-17+n -› C-14 + α

“A sneak peek of the International Conference on the Management of Spent Fuel from Power Reactors” Webinar, 3rd May 2019

thank youfor your attention