ML20042G921
| ML20042G921 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 05/14/1990 |
| From: | Stang J Office of Nuclear Reactor Regulation |
| To: | Sylvia B DETROIT EDISON CO. |
| References | |
| NUDOCS 9005160288 | |
| Download: ML20042G921 (19) | |
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j UNITED STATES JC [
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May 14,1990
. Docket No. 50-341 Mr. B. Ralph Sylvia
. Senior Vice President - Nuclear Operations Detroit Edison Company 3
6400 North Dixie Highway.
Newport, Michigan. 48166
Dear Mr.-Sylvia:
SUBJECT:
MEETING WITH DETROIT EDISON.TO DISCUSS THE RESIDUAL HEAT REMOVAL SYSTEM SMALL BORE PIPE CONNECTING FATIGUE INDUCED FAILURES On April 24, 1990, the NRC staff met with Detroit Edison Company (DECO) to discuss small bore piping failures of small bore piping on the residual heat removal (RHR) system.
A list of attendees is provided as Enclosure 1.
A copy a
of. the slides (all non proprietary) is provided in Enclosure 2.
On April 4, 1990, an initial evaluation was completed by DECO on enalytical results of vibratory stress experienced by small bore piping connection to the RHR system in the RHR pump rooms.
Vibration data obtained by Deco on the E
effected lines indicated that stresses, mainly at the welded connections on a number of lines were above the allowable ASME Section III 1971 code requirements of 12.5 Ksi.
Vibration induced fatigue failures have occurred on four small bore 3/4 inch piping connected to-the RHR system since 1987.
The failures have been attributed to pump vibration and in some cases welding flaws.
At the request of the NRC a meeting was held on April 24, 1990, to discuss the effects of the failures of the small bore piping on the operability ~of the RHR system, the ability of the RHR system to perform _its intended safety function, and for the DECO to describe the corrective action plan to resolve the small bore piping-fatigue failures.
DECO indicated that they had instituted a program to inspect or reanalysis all small bore piping connections of the RHR system.
DECO committed to repair all small bore connections with stresses greater than 20 Ksi.
The corrective actions to be taken by DECO include weldment replacement / reinforcement and/or shorting of the connections to reduce the stress concentration.
At least 11 RHR taps in the pump rooms are in the process of being modified, with work scheduled in all RHR pump rooms to be completed by the end of May, depending.
q on work duration and any LCO restrictions.
The remainder of the RHR small bore connections will be visually inspected and any needed modifications made to the connections as soon as possible, no later than the end of the second refueling cutage scheduled for March 1991.
DECO considers the RHR system is operable on the basis of the corrective actions taken.
It is DECO's assessment that any failures which could be postulated would not prohibit the RHR system from performing its intended safety function.
This is based on the designed redundancy of the RHR system and separation of divisional pumps.
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. Deco committed to submit a summarization of the finalized corrective action
'l plan, evaluations and justification for continued operation by May 31, 1990.
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Sincerely, 1
Original signed by i
John F. Stang, Project Manager j
Project Directorate III-1 Division'of Reactor Projects - III, 1
IV, V & Special Projects Office of Nuclear Reactor Regulation f
Enclosure:
As stated cc:
w/ enclosures:'
See next page
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Deco committed to submit a summarization of the finalized corrective action g.
' plan, evaluations and justification for continued operation by May 31, 1990.
Sincerely.
I i
John F. Stang, Project Manager Project Directorate III-1 a-Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation
Enclosure:
As stated cc:
w/ enclosures:
See next page
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Mr. B.. Ralph Sylvia Fermi-2' Facility
- Detroit Edison Company CC:
John Flynn, Esq.
Senior Attorney Detroit Edison Company 2000 Second Avenue:
Detroit,. Michigan 48226 Nuclear facilities and Environmental Monitoring Section Office Division'of Radiological Health P. O. Box 30195 Lansing,-Michigan 48909
.Mr. Walt Rogers U.S. Nuclear Regulatory Comission -
Resident Inspector's Office 6450 W., Dixie Highway Newport, Michigan '48166 Monroe County. Office of Civil Preparedness 963 South Raisinville Monroe, Michigan 48161' Regional Administrator, Region III
- U.S. Nuclear Regulatory Comission 799 Roosevelt Road Glen Ellyn, Illinois _ 60137 Ms. Lynne Goodman.
Supervisor Licensing Detroit Edison Company Fermi. Unit 2
- 6400- North Dixie Highway Newport, Michigan 48166-L s'
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9,1, ENCLOSURE 1
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Detroit Edison - NRC Meeting RHR Small Bore Connections April 24, 1990 9
Name Title Representive Lynne Goodman Director, Nuclear Licensing.
DECO l
. John W. Contoni Supervisor - Mechanical Engineer DECO
- Paul Fessler-Suot. Tech. Eng.
DECO t
Phillip Temple
'Prin. Engr. Weld./NDE DECO
- 1 Michael Williams Senior Engineer DECO-Stanley G. Catola V.P. Nuc. Eng. & Services DEC0 Greg Cranston Gen. Dir.-Nuc. Engineering DECO Richard Beaudry Principal Engineer DECO John A. Zwolinski.
Arst. Dir. for RG III Rx NRC Thomas 0. Martin Depty Dir. Div. of Reactor Safety, RIII NRC Bob Pierson-P/D III-1 NRC Walt.Rodgers SRI.
NRC John Stang Project Manager NRC Rober DeFayette Section Chief, DRP, REIII NPsC Mark Caruso Section Chief Reactor Sgrt Branch NRC Jake Wechsecberger Tech. Asst, for A.DIR RIII PTS NRC George Thomas Nuc. Engr.
NRR J.A. Gavula Reactor Insp. Mech.
NRC-RIII J. Rajan Mech. Eng. Br/NRR/NRC NRC L.B. Marsh Chief, Mech. Engr.' Branch NRC C.Y. Cheng Chief, Mat & Chem. Eng.-Branch NRR P.T. Kuo Section Chief, EMEB NRR B.J. Elliot Matl. Eng./DET/NRR NRR Gary M. Holahan NRC/NRR/DRSP NRR J. Partlow ADP NRR Dominic Dilanni NRC/NRR/APDIII-1 NRR t
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ENCLOSIR 2.
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DETROIT EDISON COMPANY 1:
April 24,1990 Y
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AGENDA i
e UFSAR Commitments / Codes and Standards 5
o RHR System Vibration Experience and Data jj
- Pre-op Testing / Piping Vibration Dynamics Effects Testing (PVDET) e RHR Pump Room Failure History (prior to drywell vent)
RHR Pump Room Vibration Assessment Program.
L
- Results to Date/ Fatigue Curve
- Fixes to Date s RHR System Outside of Drywell
- Inspections
- Trim Pipe Program o Engineering Functional' Analysis
- Failure Mechanism
- Acceptance Critoria
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- Site Boundary Doses e Final Fixes
- Modifications
- Schedule
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UFSAR COMMITMENTS e
t LARGE BORE SMALL BORE PIPING PIPING ASME Section til - 1971 ' Winter '71 Addenda (UFSAR 3.9) ~
For Carbon Stegl 12.5 Ksl' X
X Fig. I-9.1 @ 10 Cycles
.a 1
Stainless Steelg6 Ksl Fig.1-9.2 @ 10 Cycles
-v Piping Vibration Dynamic Effects Testing. (UFSAR 3.9) (SSER_3)
For Carbon Steel X
Sa = 7,690 pst*
l For Stainless Steel
- X Sa = 12,000 psi
[
- For Locations Selected for Monitoring (See Table 3.9-1)
Small Bore Connections X
L Visual Examination L'
FSAR E.5.110-19, Table 3.9-34 UFSAR 3.9.1 i
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RHR SMALL BORE CONNECTION FAILURES 0
. E1100 F128A o Approx.160 crack in nipple weld heat affected zone.-
-- A-Pump Discharge o Noted porosity in backside of weld following removal.'
Header Drain o NUE Weld Engineer evaluation ~- fatigue failure, outside j
surface initiated.
o 1987
' E1100-L431D' o Crack in nipple at weld heat affected zone, D Pump Suction o Fatigue failure, outside surface initiated.
Sample:Line.
o - Pump had run dead-headed.
o-1988-
'i E1100-L431C o Pin hole leak in weld close to nipple.
I C Pump Suction o NUE Weld Engineer evaluation - weld flaw.
1 Sample Line o 1989-
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0 E1100-F116A o - Approx.135 crack in nipple weld heat affected zone.
Div. I LPQ o Small Cantilevered branch with two unsupported valves.
High Point Vent o NUE Wefd Engineer evaluation - fatigue failure, outside surface initiated.
o 1990
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VIBRATION DATA RESULTS
' REPORTED CURRENT.
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STRESS STRESS MEETS "AS FOUND" AFTER ASME USING MODIFICATION ~ SEC III CONSERVATIVE (CONSERVATIVE' 1971 W 71-SIMPLIFIED SIMPLIFIED
.Sa - 12.5 PIPING DESCRIPTION MODEL COMMENTS MODEL USED)
Ksi Emergency Equipment Cooling Water to RHR Pump Shaft Seals.
- 1. Pump A'L.
Threaded-
<10,000 Yes Welded
<10,000 Yes e
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- 2. Pump B 6WM-P44-2178-1
. Threaded 19,000 Removed from 0
Yes=
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Welded
. <10,000 pump.
0 Yes
- 3. Pump C 0
6WM-P44-2179-1 Threaded
<10,000 Yes c
Welded.
- 12,500 max Yes 1,
L 4.' Pump D.
6WM-P44-2178 1 Threaded 10,100 max Yes Welded 13,400 max
. Weld Mod.
8,300 max Yes 12,800 max No y
RHR Process Line Small Bore Connections _
- 1. Taps on Pump Suction Line M
PT-E11.L409-A 12,600 Weld Mod.
7,800 Yes
-B-10,500 Yes -
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-C 14,700 Weld Mod.
9,100-Yes i;
-D N/A-Analysis l:
in Progress Weld = Mod.
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- 2. Taps on-Pump Discharge Line J
I'T-E11-L411-A 12,400 Yes
-B-11,800 Yes
-C 8,800 Yes
-D 9,400 Yes l'
3.L^ Chemical Taps on Pump Suction Lines CT-E11-L431-A 17,800 Weld Hod.
11,000 Yes
-B 10,700 Yes
--C #
18,300 Shortened
<10,000 est Yes Weld Hod.
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16,400 Weld Hod.
10,200 Yes
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Vib'-ation Data Piping Description-Taken Comments
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Pump Drain Lines 6WM-E11-2734 Line 1 Yes Line 2 Yes Non Safety Related D
-Line 3 Yes Analysis in Progress Floor Support Line 4 Yes Analysis in Progress Non Safety Related 6WM-E11-5024 Line 1 Yes Analysis in Progress Floor Support Line 2 Yes Analysis in Progress Non Safety Related Line 3 Yes Analysis in Progress Floor Support Line 4 Yes Analysis in Progress Non Safety Related 6WM-E11-5027 Line 1 Yes Analysis in Progress Floor Support Line 2 Yes Analysis in Progress Non Safety Related Line 3 Yes Analysis in Progress Floor Support Line 4 Yes Analysis in Progress Non Safety Related 6WM-E11-5029 Line 1 Yes Analysis in Progress Floor Support Line 2 Yes Analysis in Progress Non Safety Related i
Line 3 Yes Analysis in Progress Floor Support Line 4 Yes Analysis in Progress Non Safety Related j
5.
Suction Line
.i Reller Valve Piping 6WM-E11-3153 No Tied Back to Header-l
-3154 ilo Located in Torus Rm 6.
Seal Water Tubing No Stainless Steel Tubing L
No Concentrated Hasses.
7.
Piping Drains.0ff Pumps A&C Discharge Lines Trim sub. #74176
- No Pump A Weld Hod.
- 74180 No Pump C Supported 8.
Piping Drains Off Minimum Flow Lines
' Trim sub. #74022 Yes
- 71321 Yes RHR PUMP MOTOR OIL CORNECTIONS 1.
Pump 011 Drain No Short Pipe Nipple Extensions No Small (Light) valve.
- Indicates Failure
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FAILURE MECHANISM i
1 o Originates at weld toe.
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o Fatigue failures originate on outside of nipples.
o Weld discontinuities noted in examination of failures were not primarily initiators. Note:
EliU0-L431C weld pin hole was repaired prior to detailed metallurgical examination, o Failures initiate at the weld too because:
1.
Geometrical discontinuity stress riser.
2.
Metallurigical changes from weld to base metal (+ heat affected zone).
3.
Welding residual stresses also concentrate at this point.
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POST-LOCA DOSE FROM RHR-LINE BREAK-l.
PRESENT DESIGN VALUES 1.
Regulatory Limits Thyroid (rem)
Whole Body (rem) o Offsite Dose 300 25 o
Control-Room Dose 30 5
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Present Ferml' Design-Basis Dose Values (rem)
A.
Site Boundary (2 hrs) o Drywell leakage 180 6
~o 5 gpm ECCS leakage J
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Total 185 6.01 B.
Low Pop Zone (LPZ - 30 dys) o Drywell leakage 75 1.4 o
5 gpm ECCS leakage 4.2 0.3 -
0 Total 79.2 1.7 C.
Control Room o
Drywell leakage 18.7 1.5 s
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DOSE FROM RHR LINE BREAK
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Break Flow Rate (3/4" line size)-
=0 Suction line: 45 gpm o. Discharge line: 365 gpm-t 4
B.
Design Basis Scenario:. Worst break in each division.
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- t. 2 Discharge lines = 730 gpm A~
' C, Resultaiit Calculated Doses
- Assumptions j
1.
Only thyroid dose needs consideration for acceptability, 2.
.Most limiting dose occurs at site boundary.
L 3.
In no case will site-boundary thyroid dose be permitted to exceed 300 rem.
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Per the SER (NUREG-0798),5 gpm of ECCS within secondary containment equals 5 rem at site boundary. (or 1000 gallons on the floor of secondary containment equates to 8.3 rem at the site boundary) h Calculated Design Total isolation Failure Description.
Dose Basis Dose Dose Time
. o 1 RHR Pump Suction 45 rem 185 rem' 230 rem
> 2 hrs Connection o 1 RHR Pump Discharge 91 rem 185 rem 276 rem
$ 30 min.
Connections'
. o -1 RHR Pump' Discharge 110 rem 185 rem 295 rem
< 18 min.
Line in Each Division.
Loses a Connections o All Vents / Drains with 115 rem 185 rem 300 rem
< 15 min.
alternating stress >
10 Ksi fall.
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' BASIS FOR RHR-BREAK DOSES
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Standard Review Plan 15.6.5 (Appendix B): " Radiological Consequences of a Design Basis LOCA: Leakage from Engineered Safety Feature Components Outside Containment".
B.
NRC Calculations (NUREG-0798, Safety Evaluation Report) uses above methodology:
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' 5 gpm for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> --> 5 rem (SBD)
, C; Assumptions t
4 o Full classic design - base LOCA.
o 50% of core iodine into water.
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- o. Water leaks directly into secondary containment L.,
o 10% of lodine in water. released into air.
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o All iodine immediately picked up by SGTS.
o SGTS treats all lodine (99%) and releases to outside.
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ADDITIONAL DOSES AT LPZ (30 Days).
o Design. basis: 2 suction lines for 18 min.
D (total) = 79.+ 9 = 88 rem
' V.
ADDITIONAL DOSE AT CONTROL ROOM'
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- o Design - basis: 2 suction lines for 18 min.
D (total)' = - 18.7 + 2.8 = 21.5 rem i
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