ML18045A511

Forwards Semiannual Rept for Jan-June 1974 Re Leak Tests. Pages 17-18,relevant to Palisades Proceeding Encl

ML18045A511
Person / Time
Site:	Dresden, Palisades
Issue date:	08/21/1974
From:	Abel J COMMONWEALTH EDISON CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML18045A510	List: ML18045A511
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NUDOCS 8008200221
Download: ML18045A511 (3)

Text

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Common alth Edison One First N al Plaza.~icago: Illinois I Address Re 1 to: Post Office Box 767 Chicago, Illinois 60690 Enclosure l August 21, 1974 Mr *. James G. l<eppler Regional Director Directorate of Regulatory Operations - Region III U.S. Atomic Energy Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 t: .

Subject:

Dresden Station Semi-Annuai*Report,

- AEC Dkts 50~10, 50-237 and 50-249.

Dear Mr. Keppler:

Attached is the Serni-Anmral Report for Dresden Station as required by the Technical Specifications for:

Units 1, 2 and 3 for the period of January 1, 1974 through June 30, 1974.

One copy of this report* is provided for yoilr use, and 39 copies are being transmitted by this letter to the Acting.Director, Directorate of Licensing.

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Very truly yours, Nuclear Licensing Administrator Boiling Water Reactors Att.

cc.: Mr. Edson G. Case (w/att.)

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4. Surveillance ...

Al 1 1 icensed required surve 11 lance for the six month period

f. * *! January 1, 1974.'.to June _30, 1974 were satisfactorily completed.

nae unit was shutdown for refueling "operations during the entire i"* report~ng period. Required refueling surveillances were contin-

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,f... ued. lbese surveillances included such major tests as the primary containment integrated .leak rate test, core spray logic and flow tests, simulated automatic actuation test of the core spray system, station battery load discharge test, recirculation of the high pressure boron tank through the low pressure boron tank, automatic actuation test e>f th,e primary isolation valves, controi rod drive timing and latching, friction and scram tests, arid the required instrument calibrations. In addition, the inservice inspection program was satisfactorily completed as well as the other tests.

A description of the primary containment integrated leak test and inservice inspection program is given below.*

INTEGRATED PRil1ARY CONTATh."'IBNT LEAK TEST ..

A second-primary containment integrated leak rate test was

. performed during the 8th parti~l refueling. outage from January 23 through January 28. 'nle test was conducted in accordance with Technical Sp,ecification Sect ion 4. 1 .A. l .g. 'nl.is section requires a primary containment integrated leak test be conducted ac least three times in a ten (10) year period

• After the containment had been pressurized to 20 psig (Pa) on January 25, 1974, the conditions were* allowed to stabilize and leakage data was taken. After five (5) hours of data taking, the average leakage was .d.etermined to be 8.15 7./day. Investiga-tion revealed .that there was excessive leakage through an unused 2" decontamination line from the "A" cleanup system to radwaste, exiting the sphere through penetration H-45. The line was blind flanged at a spool piece connection outside the sphere in the radwaste pipeway. Additionally, leakage was noted through the secondary steam generator sample drain valve, FCV-510. Attempts to isolate this leakage failed. 'nl.e sphere pressure was then increased to 20 psig and the test resumed.

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i. After the leakage test was concluded, the verification test was

conducted in accordance with ANSI N45.4 - 1972. lbe verification l

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test lasted for approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> with the leakage calculation of .2633 +/- .034 7./day. 'nte verification test met the acceptability criteria of 10CFR.50, Appendix J which requires the verification test results to be within 0.25 L~.

Since there were two known leakage paths from the sphere during the integrated sphere test (2" decon line bl ind flange after installation and valve FCV-510) it was dee ided to do an "as found" local test on the 2" decontamination line and the secondary steam generator sample drain valve, FCV-510. After repair of the

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penetrations, a ~econ~ local test was performed and the difference subtracted from the leakage determined from the integrated leak test. The'"as fopnd" leakage t1:1rough the 2" decontamination line was found to be .0009 7./day and through the secondary steam genera*

tor sample drain, FCV-510, .019 7./day. After repairs were made to FCV-510, the calculated leakage was .0000144 7./day. After the 2" decontamination line was*seal welded, the calculated leakage was 0.00 7./day. The total difference between the "as found" and the repaired leakage is .019 7./day, thereby resulting in a primary containment integrated leakage of

• 2495 +/- *.034 7./day at a pressure of 20 psig.

The error bound of +/- 0.034 7./day is a 2<:1 value and by definition represents a 951 confidence level bound. When the primary contain-ment integrated leakage is evaluated at the upper bound, the result is 0.2835 7./day (0.2495 + 0.034). Since the 957. confidence level upper* bound leakage (0.2835 1./day) is less than the Technical Specification 4.7.A.l.e requirement (0.30 1./day) for plant startup, the test is successful.

INSERVICE"INSPECTION PROGRAM The 1973 refueling outage ins~rvice inspection program was completed on June 18, 1974 in accordance with Section XI of the ASME Code

~ (1971 Summer Addenda). During this period, approximately 220 f:

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c.omponents were examined by either volumetric, surface and/or I

visual methods. Of the indications noted, none were evaluated to be of any significance.

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' On June 18, 1974 the reactor' vessel. and associated primary system pip~ng were hydrostatically tested at 1060 psig and 280°F. At this time tw9 pin-hole leaks were detected in the same pipe section in the nort~ steam supply line to the emergency condenser. The defective pipe section was replaced according to code specifications and was satisfactorily hydrostatically tested at about 1600 psig.

5. Results of Periodic Containment Leak Rate Tests Table 5 shows the results of the periodic containment leak rate tests performed during the period from January l, 1974 to June 30, 1974.

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t 6. Changes, Tests and Experiments Reguiring Authorization From the Comnission No changes, tests, or experiments requiring co1m1Lssion authorization were performed during the period from January 1, 1974 to June 30, 1974.

.1. Key Changes in Plant Operatin~ Organization The following key changes in plant operating personnel occurred during the period from January 1, 1974 to June 30, 1974.

Station Superintendent - Ben B. Stephenson

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Operating Engineer - Eugene Budzichowski

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