ML20244E011

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Forwards Memorandum on Corrosion Evaluation - Nine Mile Point Reactor Vessel
ML20244E011
Person / Time
Site: Nine Mile Point  Constellation icon.png
Issue date: 03/04/1969
From: Cheng C
ARGONNE NATIONAL LABORATORY
To: Deyoung R
US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML17055E652 List:
References
FOIA-89-101, FOIA-89-114 NUDOCS 8904240254
Download: ML20244E011 (4)
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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 Meta 11brgy Division

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/ .; i (700 Sou:h Cass Avenue. Argonne. !!finois 00"30 To:cphone 317-739-7711

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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 s'afety on the assumption:

t f 1. the listed items of the applicant's justification are correct, I

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2. the favorab3c outcome of additional laboratory tests, and l' *

. 3. 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 ppm20 in high purity water f (c) In-service specimens in cther plants.

The table should include the following items:

j (a)' Temperature environacnt, specimen type and stress level, e

(b) Material description, heat treatment, and service treatment in 1

r t detail.

1 i '(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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3. 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. .,

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5: hTP vessel material should be set aside for future laboratory and  ;

l in-service ' r e.s t s . ,
l. 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 .viousl y acid pickled (LINO3-HF). These include:

b 1. thin valled.(0.008~to 0.028 inch) fuel cladding in solution anncaled condition..( '

2. prehenter pipes in solution annealed condition,(3) and j 6 r I

. .>. pressure vesse) liners in sensitized condition. (4,5) s t

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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 .

i and as. sensitized, plus acid pickled (HNO 3

-HF) stainicss steel pay e sily be due to the difference in the incubation period during the . initiation step. Note

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this incubation period may be shortened by corrosion fatigue.

EmEI ,0) recently reported stress corrosion cracking of sensitized 304 SS piping,

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.which had been previously mill pickled (IINO 3 M, in suHite treated water at a

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l :' 164*C, Sodiug sulfite (100 ppu)' was added to the dcmincralized water *to remove residual'o'ygen. x The oxygen content in the water analysed 0.2 ppm and the pH I

E of the water. ranged 6.5 to 7.0. Failureswereobservedinallt$rcephasesin 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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recirculating' area (uhcre the return f eed water uixes with the coolant)-

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to ovaJuate the. flowing water environment, and .

. 3. recirculation supply line adjacent to the reactor vessel to evaluate

..,[ 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) . -

N. Balai et al., Repo.t ANL-7117 Supplement (to be published)

(5)

(6) A. P. Lamick, EWL-753 (April 3 968).

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