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AHOONNd NAflONAL LABORATORY.
March 4, 1969
.t Mr.
P., C. DcYoung Deputf /.scistant j'
Dirteter of Reactor Techac]ogy Division of Reactor Licensing
.U. S. Atocic Energy Commission Uachington, D. C. 20545 Decr X. DeYouag:
Eucloscd pletse find a r.icmorandum on " Corrosion Evaluation -
Niv: Mile Tof nt Reactor Vessel".
Vc.ry truly yours, 1
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Craig Fl. Cheng g
Meta 11brgy Division E.
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(700 Sou:h Cass Avenue. Argonne. !!finois 00"30 To:cphone 317-739-7711
- TWX 910 258 3282 - WUX LB. Argenne. Ill:s c s
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Corrosion Ivalua( lon - Nine Mile Point Reactor VCSsel Although the exact mechanism of intertranular stress-assisted corrosion of sensitized Type 304 in low chloride (< 10 ppm) or chloride-frec (< 1 ppm) unter is not knoun, I believe that the ICIP plant can be operated 'with reason-ch?c assurance of 'afety on the assumption:
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the listed items of the applicant's justification are correct, I
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the favorab3c outcome of additional laboratory tests, and l'
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the installation of a material surveillance program in the reactor vessel.
Specific itcms requ4. ring submittal by the applicant crc:
1 Tahv3ated results of tests (a)70-100 ppm chloride at 180*r (b) 100 ppm 0 in high purity water 2
f (c)
In-service specimens in cther plants.
The table should include the following items:
(a)' Temperature environacnt, specimen type and stress level, j
e' (b) Material description, heat treatment, and service treatment in 1
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detail.
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'(c) Average time to failure', or no failure.
(d) Number of specimens tested or still in test.
(c)
Interrranular, transgranular, mixed failure, etc.
.(f) Expected date of test completion.
2.
Listing of estimated stress levcis and fatigue cycler for the furnacc sensitized' stainless steci components in Appendix A.
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Complete details of GE's low cycic fatigue program, and,how this
' program fac'tored into the IOfP plant, t
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4.
CD's,cva3uation of crevice corrosion and cracking of the JPDR pres-Oure vessel cladding.
s 5: hTP vessel material should be set aside for future laboratory and
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In addition I have certain reservations on applicant's conclusions. Note I
there are reported cases of intergranular corrosion cracking in reactor water 1
service with'304 stainicss steel components whose curfaces have not been pre-h,l
.v ous y acid pickled (LINO -HF).
These include:
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thin valled.(0.008~to 0.028 inch) fuel cladding in solution anncaled condition..( '
prehenter pipes in solution annealed condition,(3) and j
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I liners in sensitized condition. (4,5) t pressure vesse) s
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I The last case refers to Emm vessel cladding. - It was concluded the initial 1
cracking within the c3 adding was probably caused by thermally induced stress k
rupture during vessel f abrication. Those surface cracks propagated by low-(
cy 1c'ectrosion fatigue during reactor operatic.n. Furthermore there are two consecutive steps in the mechanism of stress corrosion cracking. The first step is ir.itiation, uhich can be triggered by' chemical, electrochemical and/or rechanical means. The second step is propagation under stress and t,he propaga-l 2 ion continues if the crack tip is not blunted or arrested. In GE's 100 ppm oxygen-ated water test, the marked difference in time to f ailure betwcen as sensitized and as. sensitized, plus acid pickled (HNO -HF) stainicss steel pay e sily be due i
3 to the difference in the incubation period during the. initiation step. Note q.
this incubation period may be shortened by corrosion fatigue.
In ad itfon,
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EmE,0) recently reported stress corrosion cracking of sensitized 304 SS piping,
.which had been previously mill pickled (IINO M, in suHite treated water at 3
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164*C, Sodiug sulfite (100 ppu)' was added to the dcmincralized water *to remove residual'o'ygen.
The oxygen content in the water analysed 0.2 ppm and the pH x
E of the water. ranged 6.5 to 7.0.
Failureswereobservedinallt$rcephasesin I
desending order of' susceptibility'y, liquid, liquf'd-vapor interface, and vapor.
I strongl'y'fcel in terms of quality assurance it is' advisable that in-scr-
- vice stressed test specimens,be inserted at the following locations
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.1.
steam space in the vessel head to evaluate oxygen and solid carry over effects,
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2.
recirculating' area (uhcre the return f eed water uixes with the coolant)-
to ovaJuate the. flowing water environment, and 3.
recirculation supply line adjacent to the reactor vessel to evaluate
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the return water environment.
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-1 (1).
W. H. Arlt and S. R. Vanderberg, Ge. Elec. Co. Report GEAP-4360 (1963).
(2)
C. N. Spaloris, Nucleonics, 21, 41 (1963).
(3) ' Combustion Eut,fnecting Co., Report CEND 265 (1965).
(4); N.' Balai et al., Report ANL-7117 (1965).
(5)
N. Balai et al., Repo.t ANL-7117 Supplement (to be published)
(6)
A. P. Lamick, EWL-753 (April 3 968).
- < 1 ppm chloride 9
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