ML20127B823

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Forwards Safety Evaluation Re Increase in once-through Steam Generator Primary to Secondary Leakage Rate,Per NUREG-1019 & Suppl 1.Rate Acceptable
ML20127B823
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 07/11/1984
From: Johnston W
Office of Nuclear Reactor Regulation
To: Lainas G
Office of Nuclear Reactor Regulation
Shared Package
ML20126B295 List: ... further results
References
FOIA-84-897, RTR-NUREG-1019 NUDOCS 8407170191
Download: ML20127B823 (5)
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o NUCLEAR REGULATORY COMMISSION h ;E WASHINGTON. D. C. 20555

.....+ W L 1 1 1984 MEMORANDUM FOR: Gus C. Lainas, Assistant Director for Operating Reactors Division of Licensing FROM: William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering

SUBJECT:

EVALUATION OF RECENT INCREASE IN OTSG PRIMARY TO SECONDARY LEAKAGE RATE AT TMI-1 In NUREG-1019 cnd Supplement No. 1 to NUREG-1019, we found the licensee's OTSG's acceptable for continued operation. By letters dated June 27, July 3, and July 6,1984, the licensee provided additional information.

This information does not affect the evaluation, conclusions or proposed license conditions in NUREG-1019 or Supplement No. 1 to NUREG-1019.

Enclosed is our evaluation.

William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering

Enclosure:

As stated cc: R. Vollmer D. Eisenhut J. Knight R. Rawson M. Wagner V. Benaroya C. McCracken J. Stolz R. Bosnak (H.%5 fiver 2 J. Van Vliet S. Pawlicki T. Sullivan P. Wu S. Kirslis L. Frank D. Sellers J. Rajan W7/7p/7/ g

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._ Safety Evaluation by The Office of Nuclear Reactor Regulation Related to Operation of Three Mile Island Unit No. 1 General Public Utilities Nuclear Corporation Docket No. 50-289

Background

In NUREG-1019 and Supplement No. 1 to NUREG-1019, we determined the once-through-steam generator (OTSG's)-had been repaired to their original licensing basis, that the applicable GOCs were met and, therefore, the OTSG's were acceptable for continued operation. By letters dated June 27 end July 3, 1984, the licensee provided additional information on i primary-to-secondary leakage rate. By letter dated July 6, 1984, the licensee provided updated information on the long' term corrosion test program.

Evaluation j

While in a cold shutdown condition, the licensee detected an increase in the background primary-to-secondary leakage rate of the B OTSG. During the

September 1983 hot functional test, the background primary-to-secondary leakage rate had been established at 1.0 gph as discussed in Supplement No. 1

. to NUREG-1019 at pg. 22. Although the increase in leakage was less than that

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' which would require inspection, the licensee opened the primary side of l - both OTSG's and conducted bubble, drip and eddy-current test (ECT) examinations.

' ECT examinations of approximately 150 tubes demonstrated that corrosion was not progressing or initiating. A total of approximately 15 tubes and/or plugs showed indications of minor leakage. Minor leakage of this magnitude is typical from plugs and anticipated from the repaired tubes. The unplugged tubes which were leaking were confirmed to be leaking between the tube and tubesheet, above the kinetic expansion repair joint. It was determined that one tube was leaking past the 6-inch repair joint sufficiently so that background leakage would have increased to an estimated 4 gph at operating I conditions. To maintain background leakage low, in addition to the leaking l

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2-tube, two additional tubes were plugged for added precaution. Leakage of this type is anticipated as discussed in various sections of NUREG-1019 and ,

- Supplement No. 1 to NUREG-1019. The total amount of primary-to-secondary leakage is limited by the plant Technical _ Specifications and the proposed License Condition No. 4 on pg. 27 of NUREG-1019, Supplement No. 1.

By letter dated July 6, 1984, the licensee updated the long term corrosion test program. Tests which simulate actual plant ' conditions with water chemistry maintained at maximum anticipated impurity concentrations continue to show no evidence of crack propagation or initiation. The absence of crack propagation or initiation in these tests continue to support the conclusion of NUREG-1019 and Supplement No. 1 to NUREG-1019.

- Test Loop No. 1 simulated conditions which assumed that peroxide cleaning would not be conducted and that a continued source of thiosulfate contamination ,

would exist. Under these conditions, one pre-existing crack in a tube specimen taken from the TMI-1 OTSG showed crack propagation. No evidence of corrosion initiation of uncracked tube specimens was detected. These test results verify the applicability of the long term corrosion test program by demonstrating that corrosion continues to propagate in water chemistry conditions where it would -

be predicted to propagate.

Conclusions l' Based on the above evaluetion, we find that the information provided by

( letters dated June 27, July 3 and July 6, 1984 does not affect the l evaluation, conclusions, or proposed license conditions in NUREG-1019 L

or Supplement No. 1 to NUREG-1019.

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.,wg Mile Island Unit t 7/16 SRI Fcx/Firo ct Inf rmation Throc mot:rcyclo riders startcd a small fire et the fcnce along 0-289 Centar the road in front of the TMI Information Center. 11.2 f i r . saas - <

quickly extinguished with no signi fi cant -dar.iage done.'

  • jgllh,',tr hroo Mile Island Unit t 9-289 7/16 SRI Fax ( On 7/16/84, the licensee notified the resident inspector i. h i t . +

five (5) steam generator (SG) plugs (3 in SG-A and 2 i n Sc. -It ) g

- I were missing from the lower (outlet) head area. Thi2 pluq loca-t i tion-verification has been performed following the cepa.r..L'uovo-J' / ment of a plug from tube 148-35 to tube 65-38 in the c ip. 4 . (in-

/ let) head area of SG-A.

The licensee's review of the tube pluqqing records indicate that-

) these plugs were in SG tubes 10-62, 133-77 and 13i-15 ( sG - /- ) . and 4

12-51 and 42-16 (SG-B) since April, 1983. Of the 1207 LG tiehc a .

plugged in the two SGs, approximately 490 have been plu:.4 al uuing i

s the Westinghouse rolled plugs.

!hrosMileIslandUnitt 7/16 SRI Fax / Seismic Qualifi- Between 9:30 a.m. and 10:00 a.m., 7/16/84, licensen repscuenta-tives determined that differential current relays fur (1.o Emar-10-289 cation of Diesel Generator Relays gency Diesel Generators (EDG's) were no t soi smically qu..l i fi ed

,' (Category I). The relays lockout the EDG's (output t.: e..L ar trip and diesel mechanical trip) on a current imbalance i s, the LI4 l

windings. The relays are manufactured by General Electsic for uso in Westinghouse switchgoar. Apparently, licenseo repre ms.lativos i

di scussions wi th these vendors confirmed that the rel.: f t. i i.n t a '. l e d

]i at THI-l were not designed for Seismic Category I usu.

j The licensee made this dotermination subsequent to t i .. s uuu..nco earlier this year of an IHP0 (Instituto of Nuclear 1:1..nt opera-4 d tions) Sincificant Event Report for a similar problo.i ideati fied at Palisades. A 10 CFR 50.72 was made at 12:43 p.o.. 1/ 1../ C i .

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$ crmant Yankee 7/16 SRI fax The licensee disclosed on 7/16/84 that RitR system in Id ;32-6

'N 50-271 was found to have an axial flaw in excess of 10% of 4.ho t hre. ugh j

wall dimensions. RHR Line 032 is part of the 20 inch ai..s *Lv.r i' suction line to the RHR pumps for shutdown coolinq apar.aticnu.

i' Weld 032-4 is located inside the drywell. The liccn.oo i.. utill ,

lj evaluating the type of repair that will be used to d i t. LI.a f l ate

' necessary. Two addi tional RHR system walds were ad.h is la tho

.ll sample of welds selected for examination.

6 i, Indian Point Unit 2 7/16 RI fax On 7/23, at 2:45 p.m., with the reactor defueled. i..am me t . ..t ipH 50-247 7/13 EMS call actuation of train A ESF occurred due to a spuriou.. c;: ; . . e . .1

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signal to the safety injection manual actuation ro l-/.

Safeguards equipment loads did not sequence onto i t .. i r.0 Volt l

busses. Operators manually reset the nocessary eqose.a s.t.. Iho

.. licensee is investigating the cause of the spuriou.. . . :.ia l .

i MORNING REPORT - REGION I i

1 DATE 7/17/84 l

j Page 3 ll.icensse/ Facility Notification / Subject Description of Items or Events

!.imtrick Information Item Mr. Chilk, Mr. Bates, and Mr. Mazuzan, will go on zin i n f.>rnat i on tour of the Limerick site with the Senior Resident i n .g..:.. t o r on i

'IS 50-352/353  !

Tuesday, July 17, 1984.

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9 MORNING REPORT - REGION V ~. p

, DATE: JULY 17, 1984

  • REGIONAL ACTION HOTIFICATION ITEM OR EVENT LCILITY

' FDLLOWUP PER MC 2515 -

)UTHERN CALIFORHIA TELEPHONE CALL / STEAM GEHERATOR TUBE LEAKS FROM RESIDENT A TUBE LEAK IN STEAM GENERATOR E-089 WAS MEASURED EDISON COMPANY INSPECTOR ON i

AT 33 GALLONS PER DAY OH JULY 16, 1984 AND AT 68 3 HGS 3 GALLONS PER DAY (0.05 GALLOHS PER MINUTE) OH' 1 50-362 7/16/84 JULY 17,1984. -THE REACTOR COOLANT GROSC ACTIVITY' IS 14 UCI/ GRAM WITH A DOSE EQUIVALAtlT 10DIHE ACTIVITY OF 0.35 UCI/ GRAM. TifE TEClit(ICAL

/ SPECIFICATION LIMIT FOR STEAM GENERATOR TUBE LEAKAGE N IS 720 GPD FOR ONE S/G OR 1 GPM TOTAL. A f.GPM TUBE I LEAK WOULD RESULT IN A DOSE RATE AT Tile SITE BoullDARY

/ OF ABOUT 0.005 MREN/HR. THE LICENSEE IS CLOSELY

/ MONITORIHG THE SITUATION AHD PLANS TO SilUT Tile UNIT

/ DOWH WELL BEFORE Tile TECH SPEC LIMIT IS REACllED.

\ Tile RESIDENT INSPECTOR IS ALSO ACTIVELY IllVOLVED

\ Ill FOLLOWING THIS NATTER.

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