ST-HL-AE-2400, Final Part 21 & Deficiency Rept Re Failure of Tubes in Component Cooling Water Hxs.Initially Reported on 871006. Discussions W/Vendor,Struthers Wells-Gulfport,Inc Indicated That Shell Side Flow Induced Vibration Problem Existed

Final Part 21 & Deficiency Rept Re Failure of Tubes in Component Cooling Water Hxs.Initially Reported on 871006. Discussions W/Vendor,Struthers Wells-Gulfport,Inc Indicated That Shell Side Flow Induced Vibration Problem Existed

ML20236L358
Person / Time
Site:	South Texas
Issue date:	11/05/1987
From:	Goldberg J HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
REF-PT21-87, REF-PT21-87-211-000 PT21-87-211, PT21-87-211-000, ST-HL-AE-2400, NUDOCS 8711100282
Download: ML20236L358 (6)

Text

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The Light Company n-, usinio e- e.a H<am

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n_,,, mis moi oi3> m.<xn November 5, 1987 j ST-HL AE-2400 l File No.: G12.408, G2.2 j 10CFR50.55(e) J U. S. Nuclear Regulatory Commission Attention: Document c'ontrol Desk Washington, DC 20555 j South Texas Project Electric Generating Station ,

Units 1 and 2 l Docket Nos. STN 50-498, STN 50-499 )

Final Report Concerning Component Cooling Water Heat Exchangers On October 6, 1987, Houston Lighting & Power Company notified your office -

pursuant to 10CFR50.55(e), of an item concerning component cooling water heat exchangers. Enclosed is our Final Report on this item. This item has been determined to be reportable pursuant to 10CFR50.55(e) and 10CFR21. y l

If you should have any questions on this matter, please contact Mr. l J. S. Phelps at (512) 972-7071. l T .

M J. H. Goldberg Group Vice President, Nuclear JSP/hg

Attachment:

Final Report Concerning Component Component Cooling Water Heat Exchangers i

B7111002B2 071105

{DR ADOCK 05000498 PDR 3

1 1 L4/NRC/JT

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i ST-HL-AE-2400 File No.: C12.408, G2.2 Page 2 cc:

Regional Administrator, Region IV Nuclear Regulatory Commission l 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 N. Prasad Kadamb!, Project Manager 'l U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814 Dan R. Carpenter Senior Resident Inspector / Operations  :

c/o U.S. Nuclear Regulatory l Commission P.O. Box 910 i Bay City, TX 77414 Claude E. Johnson Senior Resident Inspector / Construction c/o U.S. Nuclear Regulatory Commission P.O. Box 910 Bay City, TX 77414 l l

J.R. Newman, Esquire. ,

Newman & Holtzinger, P.C. l 1615 L Street, N.W.  ;

Washington, DC 20036 R.L. Range /R.P. Verret Central Power & Light Company P. O. Box 2121 Corpus Christi, TX 78403 M.B. Lee /R.P. Verret Central Power & Light Company i P.O. Box 2121 l Corpus Christi, TX 78403 i R.J. Costello/M.T. Hardt 'I City Public Service Board P.O. Box 1771 San Antonio, TX 78296 Revised 10/15/87 L4/NRC/JT.

I' Attachment ST-HL-AE 2400 j File No.: G12.408, G2.2 l Page 1 of 3 ]

1 1

i Final Report Concerning Component Component Cooling Water Heat Exchangers l l

I. Summary On October 6, 1987, HL&P notified the NRC concerning the failure of tubes l in Unit 1 and Unit 2 component cooling water (CCW) heat exchangers that '

had been determined to be potentially reportable pursuant to i 10CFR50.55(e) for Unit 2. Additional engineering evaluation was )

performed and on October 29, 1987, HL&P determined that this condition j was reportable on both Units pursuant to 10CFR21. The tube failures I resulted from shell side flow induced vibrations. Design modifications to Unit 1 and Unit 2 heat exchangers were required to prevent further damage. If the original design of the heat exchangers had remained j uncorrected, the heat exchangers' heat removal capability for-safe I shutdown or accident mitigation may not have been available.

II. Description of Deficiency On September 14, 1987 excessive makeup to the Unit 1 component cooling I water (CCW) system surge tank indicated a possible system leak. When the essential cooling water (ECW) system was isolated from the "A" Train CCW j heat exchanger, a shell to tube side rupture was confirmed. During an '

inspection of the heat exchan$er a piece of titanium tubing j (approximately 1" X 2") was found on the bottom of the west end waterbox {

(ECW outlet). Two leaking tubes were identified. The location and shape l of the tube ruptures were determined by boroscopic inspection. An air test subsequently verified that. there were no other leaks. j Subsequent to the leak identification in the Unit 1 "A". train heat exchanger, eddy current testing identified significant wall thinning in 5 other tubes in the area above the inlet impingement plate near the ruptured tubes. Similar wall thinning was identified in the symmetric I area below the inlet impingement plate. The Unit 1 "B" train heat  !

exchanger was later inspected and wall thinning was identified in the l same areas as identified in the Unit 1 "A" Train heat exchanger. 1 Subsequent to the identification of the damage on Unit 1, two (2) tube 'l ruptures in the same vicinity, were ida.ntified in the Unit-2 "C" Train heat exchanger during flushing activities (September 24, 1987). l l

Discussions with the vendor (Struthers Wells-Gulfport, Inc.) represent- i atives indicated that a shell side flow induced vibration problem existed l which would require modifications to each heat exchanger to prevent  !

further damage.

After a review of preliminary calculations by Westinghouse, performed at HL&P's request, and further discussions with Struthers_ Wells-Gulfport, Inc., a Unit 2 test program was developed to determine the magnitude of vibration and tube locations that would be susceptible to further damage. l l

I L4/NRC/JT

Attachment ST-HL-AE-2400 File No.: G12.408, G7.2 Page 2 of 3 l

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I By use of accelerometers placed in various tube locations, accelerations and displacements were measured under full flow conditions to determine which tubes could potentially be affected.

The tests identified that the number of tubes susceptible to damaging vibration (if rupture occurred simultaneously) would have resulted in lead ge exceeding the CCW System's inventory makeup ability.

III. Corrective Action The tests discussed above identified that tubes were impacting adjacent tubes and the tube support plates in the outer rows of tubes in the proximity of the impingement plate (see attached Figure 1). Based on the data taken, the pattern for tube stiffening was developed.

In each CCW heat . changer, a total of 30 tubes were removed in the areas above and below the impingement plate. An additional 264 tubes had steel i rods inserted to dampen vibration and to decrease the vibration level in I the adj acent inboard tubes (See attached Figure 1) . The vacated tube sheet holes and the rodded tubes were plugged to complete the physical modification of each heat exchanger. The plugr. used to seal the tubesheet holes and the rodded tubes are made of aluminum bronze l material, the same material as the tubesheet cladding and the ECW piping. i i

The tube and shell side film coefficients and the effective heat transfer j area (A) that make up the overall heat transfer value (UA) were affected q by the modification. A reduction in the UA vglue was offset by changing the tube side fouling factor from 0.002 Hr-Ft - F/ Btu to 0.0015. This change .1n fouling factor is based on industry standards which recommend that the fouling factor for brackish water should be 0.001. For )

conservatism, 0.0015 was used. The new UA value fog the modified heat exchanger wigh a 0.0015 fouling factor is 7.23 X 10 BTU /(hr F) compared to 6.99 X 10 BTU /(hr F). 'fhemodifgedheatexchangerintheunfouled condition has a UA value of 14.9 X 10 BTU /(hr. F). Considerable margin I still exists between cleau and fouled conditions. ]

Additionally, the ECW flow and pressure drop through the heat exchanger i tubes have been recalculated for the modified condition. The l l

calculations indicate a minimal increase in the pressure drop through the heat exchanger tubes (4.03 psid vs. 3.36 psid). The reduction in ECW flow cae to this increase in pressure drop is approximately 1.24 (184

.gpm) of the original 15374 gpm. These changes will have no impact on operability or flow balancing of' the ECW system.

L4/NRC/JT

Attachment ST-HL-AE-2400 File No.: G12.408, G2.2 l Page 3 of 3 l

l l

The seismic qualification of the heat exchanger has been re-evaluated taking into consideration the additional weight due to the modification (approximately 3900 lbs.) and has been found to be acceptable. The seismic calculation is presently being revised to include the additional weight due to the modification.

1 Other safety related heat exchangers have been reviewed to determine if they would be susceptible to the problems observed in the CCW heat exchangers. Westinghouse has developed a conservative design for their safety related heat exchangers which will support shell side flows as high as twice the design flow. The Struthers Wells heat exchangers (CCW) and Cooper-Bessemer lube oil and jacket water heat exchangers (supplied on the Standby Diesel Generators) are the only non-Westinghouse supplied-safety related heat exchangers. Heat exchangers similar to the heat exchangers supplied to STP by Cooper-Bessemer have been used on other projects with no incidence of vibration problems.

The changes that have been made co the CCW heat exchangers to resolve this deficiency do not alter any previous FSAR commitments.

Unit 1 & 2 heat exchangers have been modified.

IV. Recurrence Control I Review of the design parameters and the tests performed confirm that the i heat exchangers (after modifications) are capable of providing the heat removal capacity necessary to support the various modes of plant .

operation for both safccy-related and nonsafety-related heat loads. The l modification will prevent recurrence of this deficiency.

i V. Safety Analysis The failures in Units 1 and 2 occurred after a relatively short period of operation considering the required 40 year design life. Although only 2 l y

tubes ruptured (2 tubes each in 2 different heat exchangers), testing l indicated that the potential for tube thinning existed at other tube i locations. These other tube locations would have been susceptible to i similar failures if corrective action had not been taken. Multiple simultaneous tube failures could have resulted in a loss of component l

l cooling water (to the ECW system) greater than the CCW system make-up 4 capacity. Consequently, the CCW system may not have been kept " water soiid and as such would not have been capable of supporting safe shutdown or accident mitigation. Although the leakage would be detectable, a common mode failure could have occurred'and reduced the capability of the CCW and ECW systems to perform their safety functions.

This failure would be a substantial safety hazard. Therefore, the deficiency has been determined to be reportable pursuant to 10CFR50.55(e) and 10CFR21.

L4/NRC/JT

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