Irradiation-Assisted Stress Corrosion Cracking (presentation slides)

21 Pages·2003·0.555 MB·English

Addressing the Increasing Relevance of Irradiation
Assisted Stress Corrosion Cracking
R.Gérard
International seminar « Networking for effective R&D », Petten, 22-23 September
2003
Relevance of IASCC
Known problems:
(cid:132)
PWR: baffle bolt cracking
(cid:97)
BWR: core shroud cracking (IGSCC due to thermal
(cid:97)
sensitization with possible assistance of radiation effect?)
Potential problems:
(cid:132)
PWR: baffle bolts/plate swelling
(cid:97)
PWR: long term evolution of IASCC susceptibility
(cid:97)
Maximum neutron dose (approx.) - 40 years of operation
PWR (typical 900 MW unit)
(cid:132)
Core barrel ≈ 10 dpa, T≈ 300 – 330°C
(cid:97)
Core baffle ≈ 80 to 100 dpa T≈ 370°C (at former levels)
(cid:97)
Baffle bolt ≈ 70 dpa, T ≈ 350°C
(cid:97)
BWR (typical)
(cid:132)
Shroud and top guide ≈ 0.6 dpa
(cid:97)
Fuel support ≈ 7 dpa
(cid:97)
PWR lower internals (Typical)
PWR core barrel
Baffle-formers
assembly
Baffle/formers/core barrel assembly (detail)
CW 316 SS 5/8” bolts
With or without cooling holes
Baffle bolt cracking
Cracked baffle bolts detected in French, Belgian, Japanese
(cid:132)
and U.S. plants (316 CW and 347 stainless steels)
Examination of extracted bolts confirm IASCC origin of the
(cid:132)
cracking
Number of bolts largely exceeds what is necessary to
(cid:132)
maintain the integrity of the internals in accident conditions
(control rods insertion)(cid:206) no safety issue as long as
cracking remains limited
No reliable predictive model to date; influence of dose and
(cid:132)
stress
Baffle bolt cracking evolution in French CP0 units and
Tihange 1
100
Fessenheim 1/Tihange 1 *
* The bolts of the plants
90 Fessenheim 2 / Bugey 2 * represented by the same
Bugey3 symbol are made
TS 80 Bugey4 from the same heat
L
O Bugey5
B
D 70
E
K
C
A Bugey 2
R 60
C
D
E
M 50 Fessenheim 2
R
FI
N
O 40
C
F
O
R 30
E Tihange 1
B
M
NU 20 Bugey 3
Fessenheim 1
10 Bugey 5
Bugey 4
0
50000 65000 80000 95000 110000 125000 140000
EQUIVALENT FULL POWER HOURS
From “An analysis of Bafle/Former bolt cracking in Franch PWR” Peter M. Scott, M.C.Meunier, D.Deydier, S.Sylvestre and A.Trenty,
Environmentally Assisted Cracking : Predictive methods for Risk Assessment and Evaluation of Materials , ASTM 1401 “
Irradiation swelling
Problem known from fast reactors experience
(cid:132)
Until recently, not expected to occur at PWR temperatures
(cid:132)
First measurement on non-negligible swelling reported in
(cid:132)
Tihange 1 Baffle Bolt in 2000 (CIR): 0.25% max
Lower levels measured in baffle bolts extracted from
(cid:132)
U.S.plants (0.03%)
Irradiation swelling is highly temperature dependant
(cid:132)