ML20210P849
| ML20210P849 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/14/1997 |
| From: | Stanley H COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20210P853 | List: |
| References | |
| GL-95-05, GL-95-5, NUDOCS 9708270376 | |
| Download: ML20210P849 (4) | |
Text
o.mmmm ranh hi.on rumlun>
i tirahlwia nd t,rstr ting st: tion j
nome. Im 8 e Drat rulir.11. (d007 %19 lot His 64H 2801 August 14,1997 Document Control Desk Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
Ilraidwood Station Unit i Steam Generator Interim Plugging Criteria 90 Day Repon NP37 72; NRC Docket No. 50-456
References:
I hiny 14,1997 letter from h1 D. Lynch, Oflice of Nuclear Reactor Regulation to I. Johnson, issuance of Amendments (TAC NOS.
h196498, h196499, h196500 AND h196501) 2.
NRC Generic Letter 95 05, Voltage Based Repair Criteria for Westinghouse Steam Generator Tubes AfTected by Outside Diameter Stress Corrosion Cracking.
3 hiny 29,1997 letter from T. Tulon to U.S. Nuclear Regulatory Commission, tiraidwood Station Unit 1 Cycle 6 Refuel Outage Steam Generator Inservice inspection Report, Docket No. STN 50 456
- 4. hiceting between NRR and Comed on July 23,1997 Reference 2 requires the results of a Steam Generator (SG) tube eddy current inspection be issued to the stair within 90 days following plant restart. Restan (hiode 2)liam the Braidwcod Unit I sixth refuel outage (AIR 06) was on hiay 25, 1997. Pursuant to these requirements, Comed is submitting th enclosed report from the Braidwood Unit i AIR 06 steam generator inspection. A summary of the inspection results and evaluations required by References I and 2 is as follows:
\\
- 1. The bounding Probability of Burst (POB) for the End of Cycle 7 (EOC-7) was Q
ibund to be in the l A SG (8.0x10 ). The 1 A SG POB is less than the lx10'2 t\\
4 limit specilled in Reference 2
- 2. The bounding hiain Steam Line Break (hiSLB) leak rate at the EOC-7 was
(\\
found to be in ihe IC SG (57.1 gallons per minute at Room 9700270376 970814 IIM N N:WllNtuas PDR ADOCK 05000456 15,5,E m.lu su.llll.lilil.l p
pg
,, si e N kIDkOttk k410l Af y
U.S Nuclear Regulatory Coniniission I
Document Control Desk August 14,1997 l
in 100% of the tubes inspected. No indications detrimental to the integrity of the load path necessary to support the 3.0 volt IPC were identified.
- 11. The as.found EOC 6 MSLil leak rate was found greater than the projected i
EOC 6 leak rate. An investigation was conducted to identify the root cause of this underprediction A voltage dependent growth rate was determined to be the cause of the underprediction. The probe wear criteria was determined to have an insignificant contribution to the underprediction. The EOC leak rates l
reported in the attachment include application of the voltage dependent growth rates.
Attachment A to this letter contains the liraidwoed Unit i Cycle 7 Interim Plugging Criteria 90 Day Report.
As discussed in Reference 4 the site allowable leak rate values and the predicted leak rate values have not been reported under the same conditions.
For liraidwood Unit 1, in the previous 90 Day Reports, the site allowable MSLil leak rate limit has been reported to be 9,4 gallons per minute (gpm), including 0.1 gpm from each of the unfaulted SGs, with a reactor coolant dose equivalent iodine 131 limit of 1.0 microcuries per gram The bounding MSLB leak rate in the 90 Day Report is compared against this site allowable limit. While preparing the current 90 Day Report,it was determined that the 9.4 gpm value was measured at reactoi coolant temperature and pressure. The MSLB leak rate, calculated in the 90 Day Report, is at Room temperature and pressure. The difference between the water density at reactor coolant temperature / pressure and Room temperature / pressure is
~
a factor of 1.406 in the non conservative direction. As a result of this, the predicted and actual MSLil leak rates calculated in the past 90 Day Reports were reviewed to determine if the site allowable limit had been exceeded. The table below lists the predicted and actual MSLB leak rates along with the site allowable limit for each of the past outages where an interim plugging criteria was
' implemented Cycle Bounding Sounding Site Allowable Site Allowable Predicted SG Actual SG IPC Leak Rate Leak Rate IPC Leak Rate Leak Rate Limit' Limit *
(gpm @ Room (gpm @ Room (gpm @ RCS (gpm @ Room Temp / Press)
Temp / Press)
Temp / Press)
Temp / Press) 4 N/A 1.71 9.4 6.7 SA 2.81 0.32 26.8 19.0 58 0.48 0.07 26.8 19.0
~
6 6.99 11.5 26.8 19.0 7
57.1 N/A 26.8" 19.0"
- Includes 0.1 gpm from each of the three unfaulted SGs
" A Technical Specification amendment is being requested to lower the RCS DE l 131 limit to bound the predicted leak rate, i
e
-G--..~;.-.
~. _.,
_... - - _. - - -. ~ -, - -. - -.
U.S. Nuclear Regul: tory Commission Document Control Desk August 14,1997 Temperature / Pressure). Operation for 144 days at an operating temperatuie greater than $00T is allowed under the current reactor coolant dose equivalent iodine 131 limit of 0.35 microCuries per gram.13raidwood Station is preparing a Technical Speci0 cation amendment request to sufliciently reduce the Unit I reactor coolant dose equivalent iodine 131 limit to an acceptable level for full cycle operation.
- 3. One circumferential crack like indication was detected at a tube support plate intersection. This intersection contained a large dent. This indication was removed from service. Circumferential crack like indications were also detected in the tube roll transition region at the top of the tubesheet in tubes that were locked to support the 3.0 Volt Interim Plugging Criteria.
A Westinghouse Laser Welded Sleeve was installed in each of the tubes that are locked to support the 3.0 Volt laterim Plugging Criteria.
- 4. No Primary Water Stress Corrosion Cracking (PWSCC) was detected at TSP intersections. No interference from copper deposits was identined.
- 5. No indications were found to extend beyond the confines of the tube support plates.
- 6. No corrosion induced dents were identified in tubes adjacent to the tube intersections that were expanded at the Tube Support Plates to support the 3.0 Volt Interim Plugging Caiteria.
- 7. A 0.610 inch bobbin coil probe was used to inspect the TSP intersections except for the hot leg portion of six tubes. These tubes were in the first three rows of tubes and had a Westinghouse Laser Welded Sleeve installed at the top of the tubesheet on the hot leg side. The sleeve prevented the use of a 0.610 inch probe. The tight U ilend radius prevented inspecting the hot-leg portian of the tube from the cold leg side.
IPC was not applied to the TSP intersections inspected by the smaller probe.
- 8. The Upper Voltage Repair Limit only applies to the Cold Leg indications. The Upper Voltage Repair Limit was 2.04 volts.
There were no Cold Leg indications greater than the Upper Voltage Repair Limit. There were two Cold Leg indication greater than 1 volt and corifirmed by a Rotating Pancake Coil probe.
- 9. Two tubes were removed from the Steam Generators. The metallurgical results of these pulled tubes are included in the attached report.
- 10. Eddy current analyses were performed on tubes near the Anti Rotation Devices and around a Patch Plate seam to monitor for degradation of the Tube Support Plates. Also, the presence of a Tube Support Plate signal was verified
U.S. Nuclear Regulatory Coniniinion Document Control Desk August 14,1997 The result of this review showed that, although the site allawable leak rate limit was less than previously believed, none of the cycles exceeded the limit.
I' lease direct any questions regarding ihis submitta', to Terrence Simpkin, tiraidwood 1(egulatory Assuiance Supervisor,(815) 458 2801, extension 2980.
Very truly yours, l
=
Il Gene Stanley Site Vice l' reside t liraidwood Nuclear Generating Station n.,w..
Attachment ec:
C. Phillips, Senior itesident inspector liraidwood G. Dick, Iltaidwood Project Manager - Nitf(
A,11. Ileach, Regional Administrator Rlli Ollice of Nuclear Safety. IDNS