ML19326C395
| ML19326C395 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/16/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Phillips J ARKANSAS POWER & LIGHT CO. |
| References | |
| NUDOCS 8004220940 | |
| Download: ML19326C395 (5) | |
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DISTRIBUTION:
JMMcGough, DRL Docket-4 5
VHWilson, DRL (2) AEC PDR HRDenton, DRL Local PDR RWKlecker, DRL Docket RRMaccary, DRS DRL Reading PWR-3 Reading EGCase, DRS Att"}*7 Docket No. 50-313 3y DR SHHanauer, DR FEB 161972 DJSkovholt, DRL PFCollins, DRL Mr. J. D. Phillips R
ung, DRL Vice President and daief Engineer FSchroeder, DRL Arkansas Power Em Light Company TRWilson, DRL Sixth and Pine Streets PWR Branch Chiefs Pine Bluff. Arkansas 71601
Dear Mr. Phillips:
As you know, an event occurred at a foreign pressurized water power reactor in which an unusual corrosion mechanism occurred when prolonged leakage of borated reactor coolant onto the reactor vessel head was undetected. Subse-quant tests have indicated that this corrosion potential s.ight exist under certain conditions when borated fluid has prolonged contact with carbon steel.
To preclude additional experiences of this type, an appropriate program of inservice inspection should be implemented to detect such effects at an early s tage. The ASME Code canaf ttee for Inservice Inspection is considering revision of the ASME Code for Inservice Inspection of Nuclear Reactors.
However, as an interim zeesure, we believe that the inspection program described in the enclosure should be incorporated into your inservice inspection program.
Please advise us within thirty bays concerning your adoption of the pro-visioct of the enclosure.
Sincerely, Original Signed by R. C. DoYoung R. C. DeYoung, Aseistant Director for Pressurized Water Reactors Division of Reactor Licanaing
Enclosure:
PWR Inservice Inspection Program ec w/ encl: See attached l
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Foran AEC-518 (Rev. 9-53) AECM 0240
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2 FEB 161972 Mr. J. D. Phillipe l
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Mr. Horace Jewell House, Holas, & Jewell 1550 Tower Building Little Rock, Arkansas 72201 Mr. Roy 3. Snapp 1725 K Street, N. W.
Washington, D. C.
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DATE >................
Form AEC-SIS (Rev.9-33) APCM 0240 U. s, covenmasENT rnorreso orrics : 1sto o. 40s-ses
Recommended PWR Inservice Inspection Program for Detection of Effects of Reactor Coolant Leakage A.
Inspection Requirements (1) Prior to reactor startup following each refueling outage, all pressure-retaining components of the reactor coolant pressure boundary shall be visually awa=fnad for evidence of reactor coolant leakage while the system is under a test pressure not less than the nominal system operating pressure at rated power.
This examination (which need not require removal of insulation) shall be performed by irspecting (a) the exposed surfaces and joints of insulations, and (b) the floor areas (or equipment) directly underneath these components.
At locations where reactor coolant leakage is normally expected and collected (e.g., valve stems, etc.), the examination shall verify that the leakage collection system is operative and leaktight.
(2) During the conduct of the examinations of (1) above, particular attention shall be given to the insulated areas of components constructed of ferritic steels to detect evidence of boric acid residues resulting from reactor coolant leakage which might have accumulated during the service period preceding the refueling outage.
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. (1) The vtaual examinations of (1) and (2) above shall be conducted in conformance with the procedures of Article IS-211 of Section XI of the ASME Boiler and Pressure Vessel Code.
B.
Corrective Measures (J) The source of any reactor coolant leakage detected by the examina-4 tions of A(1) above shall be located by the removal of insulation i
where necessary and the following corrective measures applied:
1 (a) Normally expected leakage from component parts (e.g., valve stems) shall be minimized by appropriate repairs and mainten-ance procedures. Where such leakage may reach the surface of ferritic components of the reactor coolant pressure boundary, the leakage shall be suitably channeled for collection and disposal.
(b) Leakage from through-wall flaws in the pressure-retaining mem-brane of a component shall be eliminated, either by corrective repairs or by component replacement. Such repairs shall con-form with the requirements of Article IS-400 of Section XI of the ASME Boiler and Pressure Vessel Code.
(2) In the event boric acid residues are detected by the examinations of A(2) above, insulation from ferritic steel components shall be removed to the extent necessary for examination of the component
3 surfaces wetted by reactor coolant leakage to detect evidence of corrosion.
The following corrective measures shall be applied (a) An evaluation of en sffect of any corroded area upon the structural integri* of the component shall be performed in accordance witn the provisions of_ Article IS-311 of l
Section il Cod-(b) Repairs cc roded areas, if necessary, shall be perferneo in accordance with the procedures of Article IS-400 of Section XI Code.
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