ML19309G366
| ML19309G366 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 04/28/1980 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML19309G367 | List:
|
| References | |
| LAC-6882, NUDOCS 8005060220 | |
| Download: ML19309G366 (6) | |
Text
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D DA/RYLAND aoo o y2 h
COOPERAT/W eo oox ai7 2615 EAST AV SOUTH
- L OSSE.W CO S N M601 (608) 788-2000 April 28, 1980 In reply, please refer to LAC-6882 DOCKET NO. 50-409 Mr. James G.
Keppler Regional Director U.
S. Nuclear Regulatory Commission Directorate of Regulatory Operations Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR)
PROVISIONAL OPERATING LICENSE NO. DPR-45 REPORTABLE OCCURRENCES NOS. 80-02 AND 80-03
References:
(1)
LACBWR Technical Specifications, Section 3. 9. 2.a. (3).
(2)
LACBWR Technical Specifications, Section 5.2.1.2.(b)4 and 5.2.1.3.
(3)
DPC Letter, Linder to Keppler, (LAC-6865), dated April 16, 1980.
(4)
DPC Letter, Linder to Keppler, (LAC-6869), dated April 17, 1980.
Dear Mr. Keppler:
In accordance with Reference 1, this submittal constitutes the required followup report describing the occurrences which were initially reported in References 3 and 4.
The subject occurrences involved the discoveries of degradation in the primary reactor containment during periodic Type C surveillance leakage rate testing whi'o in a shutdown condition.
As a result of conducting periodic Type C leak tests on Containment Building ventilation dampers on April 15, 1980, the leakage flow rate determined from the exhaust dampers exceeded the permissible leakage flow rates as defined in Reference 2.
The test leakage rate was determined to be 40.79 SCFH.
The acceptance criteria is 0.375 SCFH.*
Supplemental tests verified zero leakage across the contain-ment boundary.
The method of leakage rate determination used for this test requires pressurizing the ventilation piping space between the two dampers
- Shall not exceed 1% of leakage rate L where L is 37.5 SCPH.
Po, pg Mr. James G. Keppler, Reg. Dir.
LAC-6882 U. S. Nuclear Regulatory Commission April 28, 1980 with air to 52 psig.
After isolating the air supply, pressure decay versus elapsed time is observed.
(See attached Figure 1).
As a means to assist in the determination of the point (s) of leakage, a blank flange was installed downstream of the downstream damper (73-25-006) and flow measurements made.
No leakage could be detected past the lower dampers, showing that containment integrity had been maintained.
Further investigation revealed the source of leakage to be the seat ring of the upstream damper.
An indentation, approximately 1/8" wide and 1/16" deep, located on the surface of the seat ring caused the leak.
The indentation was formed by contact of the rim of the valve disc when the valve disc is in the open position.
A similar indenta-tion on the opposite side of the disc did not leak The ventilation outlet damper (73-25-005) is an Allis-Chalmers Stream Seal No. 150-R, 10-in. butterfly valve.
(See Figures 2 and 3).
Indentations in the seat ring have been observed before.
The upstream and downstream exhaust valves are installed with the same damper orientation.
The tests results show that when the test pressure is applied against the seat ring and damper in the direction which would be experienced during an MCA, the valve is leak tight.
The seat ring was replaced in the upstream damper and both dampers were repacked with the square type braided teflon packing first used in May, 1979.
Use of this new type of packing was discussed in the report on Reportable Occurrence No. 79-08 (LAC-6322, dated May 30, 1979).
Experience to date with this packing has been good, as evidenced by no leakage occurring past the lower damper operator shaft during this test.
A post-repair leakage test performed April 18, 1980, resulted in an indicated leakage rate of 0.05439 SCFH, in conformance with the Technical Specification acceptance criteria.
As a result of conducting periodic Type C leak tests on the Contain-ment Building Condensate Return Isolation Valve on April 17, 1980, j
the leakage flow rate determined from the condensate valve exceeded the permissible leakage flow rates as defined in Reference 2.
The test leakage rate was determined to be 3.6 SCFH.
The acceptance criteria is 0.375 SCFH.*
1 The Containment Building Condensate Return Isolation Val *ra (73-25-021) is a Fischer, Model 667ET, 1 " valve, with r carbon impregnated teflon seal ring.
This was the first failure of thia valve and can be attributed to aging due to normal wear.
The talve
- Shall not exceed 1% of leakage rate L where L is 37.5 SCFH.
po, po 9
Mr. James G. Keppler, Reg. Dir.
LAC-6882 U.
S. Nuclear Regulatory Commission April 28, 1980 seals were replaced and the post-repair leakas test performed April 25, 1980, resulted in zero leakage.
No further corrective action is considered necessary.
Licensee Event Reports (
Reference:
Appendix A, Regulatory Guide 1.16, Rev. 4) for Reportable Occurrences No. 80-02 and 80-03 are enclosed.*
Should you have any questions regarding this submittal, please contact us.
Very truly yours, DAIRYLAND POWER COOPERATIVE Frank Linder, General Manager FL:LSG:af Attachments cc:
Director, Office of Inspection and Enforcement (30)
U.
S. Nuclear Regulatory Commission Washington, D. C.
20555 Director, Office of Management Information (3) i and Program Control U.
S.
Nuclear Regulatory Commission Washington, D. C.
20555
- Authorization for this report to be submitted beyond the fourteen-day reporting period was granted to L. Goodman by Mr. Ken Ridgway on Tuesday, April 29, 1980.
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