ML20127G538
| ML20127G538 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/25/1973 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9211170360 | |
| Download: ML20127G538 (2) | |
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to Mr. A. Giambusso Deputy Director for Peactor Projects yC United States Atomic Energy Commission Washington, D C 20545
Dear Mr. Gi ambusso:
f.OfRICELLO f1JCLEAR GENERATIf6 PLAta Docket tb. 50-263 License No. DPR-22 Primary Containmont isolation Valve Leakage A condition occurred at the Monticello Nuclear Generating Plant recently which we are reporting to your office in accordance wi th provisions of Section 6.6.B.3 of Appendix A, Technical Specifications, of the Provisional Operating License DPR-22.
Three primary containment isolation valves were found to have seat leakape creater than the Technical Specification limit of 17.2 scfh at 41 psic. 'The leaking valves were A0-2378 (Suppression chamber purge inlet valveh A0-2390 (Reactor building to suppression chamber vacuum breaker valve), and AD;2383 (Suppression chamber purce outlet valve).
These valves are 18" and 20" of fset disc butterfly valves with a "T" chaped inflatable seal for positive seat
- sealing, in each case a redundant primary containment isolation valve on the same line was found to be within the Technical Speci fication limi t.
On December 19, 1972, while pressurizing the drywell through the 18" air purge supply line for the purpose of leak testing the torus to drywell vacuum breaker valves, personnel inside the torus detected leakage through A0-2378.
A local leak rate test was attempted on the section of the piping bounded by A0-2378, A0-2381 (the drywell purge inlet valve) and A0-2377 (the purge line isolation valve outboard of A0-2378 and A0-2391).
Due to the extent of the leakace, the piping section could not be pressurized to the local leak test pressure required by the Technical Speci fications (41 psig).
A0-2378 was cycled several times but proper disc sealing could not be obtained.
I t was postulated that the excessive leakage might be due to either improper adjustment of the valve linkage or degradation of the resilient seal.
The valve actuator to stem linkage was adjusted to position the valve disc on a new area of the seal.
A subsequent leak rate test indicated a total leakage of 17.14 scfh past the three valves.
9211170360 730125 PDR ADOCK 05000263 S
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As a result of the A0-237R leakge, oil similar valves in the plant were tested by pressurizing settions of piping between isolation valves, lhese tesis, which were cocpleted on January 17, 1973, indicated excessive leakare from the sections of piping containine valves A0-23?O and AD-2383.
Adjusinent of the actuator to stem linkare on A0-2383 decreased the reasured leakage to 9.36 scf h.
In addi tion to linkage adjusinent, i t was necessary to exercise the seal on AD-2390 by al ter-nating aplications of pressure and vacuum to obtain acceptable leakage.
lo assure that the ceal was truly operating f reely, A0-2380 was cycled 25 tires and then leak te d ed wi th acceptable results af ter 3 additional cycles.
T he final leakape measurenent was 7.24 scf h.
The problem will be investigated fur ther during the refueling outage planned fo r f/ arch,1973 and the results of this investication will dictate what further action is required on these three valves as well as similarly constructed valves.
At present, all suppression chanber and drywell purge inlet and outlet valves and the reactor building to suppression chamber valves are closed and sealed properly.
To veri fy that this sealing capability is maintained between now and the refueling outage, a leak rate test will be performed following each cycling of these valves.
In the initial integrated leak rate test of the primary containcent, the total leakage was determined to be 0.437 weight percent of the contained air during a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
Taking the conservative approach that the present leak rates are attributable to the primary containment valve on each line that was not adjusted, and addinc 17.14 scfh, 9.36 scfh, and 7.24 scfh to the previous total leakage, the total leakage becomes. 537 weight percen t.
This leakage is within the Technical Speci fication limi t of 1.2 weight percent.
An Abnormal Occurrence report will be available at the si te for the Regulatory Operaiions inspoctor.
Very truly yours,
. 0. /
L. O. Mayer, PE Di rector of Nuclear Support Services l.0 Wk i k cc:
B d Grier G
Charnoff flinnesota Pollution Control Agency, Attn:
K. DzuEan
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