ML20127B080

Board Notification 83-184:provides Addl Info on Steam Generator Repair Program & Requests Concurrence on Final No Significant Hazards Determination Re Operation W/Repaired Generators.Commission Papers Encl

ML20127B080
Person / Time
Site:	Crane
Issue date:	11/23/1983
From:	Lainas G Office of Nuclear Reactor Regulation
To:	Atomic Safety and Licensing Board Panel
Shared Package
ML20126B295	List: 05000289/LER-1981-013, Forwards LER 81-013/01T-0 Re Leaking Found in Steam Generators.Event Reportable Per Tech Spec 6.9.2.A(3). Supporting Documentation Encl IA-84-897, Forwards Safety Evaluation Re Increase in once-through Steam Generator Primary to Secondary Leakage Rate,Per NUREG-1019 & Suppl 1.Rate Acceptable ML20126B293 ML20126B302 ML20126B316 ML20126B321 ML20126B337 ML20126B342 ML20126B358 ML20126B376 ML20126B390 ML20126B412 ML20126B435 ML20126B467 ML20126B475 ML20126B493 ML20126B504 ML20126B513 ML20126B520 ML20126B560 ML20126B563 ML20126B585 ML20126B592 ML20126B610 ML20126B618 ML20126B623 ML20126B635 ML20126B639 ML20126B648 ML20126B650 ML20126B656 ML20126B658 ML20126B667 ML20126B670 ML20126B684 ML20126B688 ML20126B701 ML20126B708 ML20126B716 ML20126B719 ML20126B739 ML20126B751 ML20126B758 ML20126B766 ML20126B784 ML20126B797 ML20126B806 ML20126B825 ML20126B831 ML20126B879 ... further results
References
FOIA-84-897, TASK-AS, TASK-BN83-184 BN-83-184, NUDOCS 8312140274
Download: ML20127B080 (18)

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UNITED STATES r-

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\\g jtlV Noverrber 23, 1983 Docket No.

50-289 MEMORANDUM FOR:

Atomic Safety and Licensing Board for TMI-1 Steam Generator Repair FROM:

Gus C. Lainas, Assistant Director for Operating Reactors, DL

SUBJECT:

BOARD NOTIFICATION (BN-83-184 ) TRANSMITTING COMMISSION PAPER ON TMI-1 STEAM GENERATOR REPAIR

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In accordance with the NRC procedure for Board Notification, the following information is being provided for your information.

On November 18, 1983, the staff forwarded to the Commission the subject paper (enclosed) to provide the Commission with additional irformation on the TMI-1 steam generator repair program and to request Commission concurrence on the final no significant hazards consideration determination regarding operation with the repaired steam generators.

Information transmitted to the Commission included the final determination, Supplement 1 to NUREG-10.19, and a draft license amendment which would permit return to operation of the repaired steam generators (all enclosed).

Pending Commission action, the license amendment-has not been issued.

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/'Gus C. Lainas, Assistant Director A[

for Operating Reactors, DL A et

Enclosure:

Commission paper with geb('

Enclosures lc 'l cc: See next page

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' FINAL DETERMINATION OF NO SIGNIFICANT-HAZARDS CONSIDERATION' RELATED TO AMENDMENT NO.

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-FACILITY OPERATING LICENSE NO. DPR GPU NUCLEAR CORPORATION, ET AL THREE MILE ISLAND NUCLEAR STATION,' UNIT 1 (TMI-1)

DOCKET N0. 50-289 OPERATION WITH REPAIRED STEAM GENERATORS r

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h The OTSG hot functional test program was to take approximately thirty

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days and include extensive leak testing and transients which will maximize g

stresses on the tubing.__---

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/ By i~etter dated October 25, 1983, the licensee provided TDR-488 which

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contains information on the OTSG pre-critical (non-nuclear) hot func-

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h tional testing.

Primary-to-secondary leakage was monitored during the /

5; entire test period, using Krypton-85 monitoring as discussed in i

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i Section 3.5 of this supplement.

BasedontheOTSGleakageresults,thel I

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-licensee has established a baseline leakage rate of 1.0 GPH which is tC

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'r 1/60 of the Technical Specification limit.

This low rate of primary-.j M,

to-secondary leakage provides additional confirmation that the kinetic ]

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. expansion procedure is an e.)lective repair method.

During subsequent

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operations, if leakage inc'reases by 0.1 GPM (6.0 GPH) above background "Ih the plant will be shut down, the OTSG's examined and repaired as

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necessary.

Reactor coolant system analysis showe,d sulfate concentra-iik tions between 20 ppb and 76 ppb, which provides additional confirmation i

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td that the principal sources of sulfur have been removed.

The low base-ft; line primary-to-secondary -leakage and low reactor coolant sulfate

,y PP concentrations confirm our conclusions in the SER.

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3 3.8 Occuoational Dose Assessment Fp

d In the SER, we discussed the occupational dose received during various itT portions of the tube repair program as well as that required to complete the j

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Topical Report 008 Revision 3 presented revised figures for the

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completed program.

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INTRODUCTION

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On August 25, 1983, the Nuclear Regulatory Commis i its Safety Evaluation dealing with the steam ge s on issued NUREG-1019, return to operation of TMI-1.

nerator tube repair and repair program was acceptable, that appropriate GThe y

yy (GDC) had been met, and that subject to resoluti eneral Desi

k identified in Section 5.3, there is reasonable as on of open items and safety of the public will not be endangered by o surance that the health with the repaired steam generators.

peration of TMI-1 4 :

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Since issuance of NUREG-1019, the licensee has p in Revision 3 to its Topical Report 008 and in itrovided additiona pr which included Revision 2 of TDP.-406 (SGTR Guidelis letter of b

I 30, 1983, on NUREG-1019.

1.eak/ Rupture)) and Emergency Plan Imp 1

G Tube Dose Projections)'have been made available to th r ce ure 1004.7 5

letter of October 25, 1983, e staff.

Testing Results and Evaluation.the licensee submitt d TD In addition, by R-488, TMI-1 OTSG Hot e

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The purpose of this Supplement is to update NUREG 10 above information.

19 by addressing the In the SER, we provided a description of the re on the 22-inch kinetic expansions which are limiting in dpair me tube pullout from the tubesheet cannot occur etermining that conditions as a consequence of severance of the tubes at thu transition zone.

By letter dated September e tube repair that our SER did not clearly indicate that tube 30, 1983, the licensee noted 22-inch and 17-inch kinetic expansions.

s were repaired using both We agree with the licensee's coments

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f The majority of tubes were repaired using a 17-inch expansion because the vast majority of defects were located near the top of the upper tubesheet.

The 17-inch expansions provided for repair of tubes with defects down to g

11 inches from the top of the tubesheet while retaining the 6-inch Tubes with defects between 11 inches and 16 inches 3

qualification zone.

were repaired using 22-inch expansions.

Because the 22-inch expansion, d

which is the limiting case, was already addressed, the information does y

E not alter the conclusions in our SER.

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3' EVALUATION OF REPAIR 5

h 3.1 Determination of Causative Acent(s)

-lA In the SER, at the top of page 8, we stated that "The thiosulfate tanks y

have also been physically removed."

By letter dated September 30, 1983, the licensee pointed out that the lines which connect the thiosulfate tank g

lj to the reactor coolant system have been physically severed and sealed but h

This information does not alter our conclusion 5

the tank has not been removed.

in the SER.

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t Examination and Repair of the Remainder of the RCS k

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In the SER on page 9, we stated that "all corrosion-affected sections in the

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waste gas system have been replaced."

By letter dated September 30, 1983, ble

.the licensee noted that only sections of the waste gas system with unaccep

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corrosion have been replaced.

Piping with minor corrosion indications will be placed on an augmented inspection list. We agree with the licensee, because

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sections of this low pressure system in which the corrosion indications were not significant need not be replaced.

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. The third paragraph on page 9 of _ the SER should now read:

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~"During a supplemental examination of systems which interface with the reactor coolant system, evidence of sulfur-induced corrosion was found tin a~ waste gas _ system stainless steel line The extent of corrosion was l quantified and all sections with identified cracks have been replaced

. One l valve in the waste gas system exhibited an indication which could not be identified as a defect.

This valve has been placed on-the in-service inspection list for further monitoring.

In addition, the power operated relief valve (PORV) was' removed for examination.

Components of the PORV

.were found to exhibit pitting corrosion attributable to sulfur which could-have reduced.the yalve's capability to function but did n' t affect o

its. structural integri'ty. At a meeting on May 20, 1983, the licensee pro'vided results of the pressurizer corrosion examination.

Examination of the -PORV block valve, the connecting piping, safety relief valves, and the remainder of the pressurizer revealed only shallow pitting on a non-seating surface of one.of the two safety relief valves.

Based on this examination, the PORY was rebuilt with uncorroded parts.

Both safety relief valves were replaced for reasons other than corrosion.

?The conclusion on page 13 of the SER-should now read:

Conclusion The staff finds 'that the' PORV was refurbished with uncorroded parts.

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Additionally, the staff notes that, although some light pitting was found on'one of the two. safety valves,~the pitting was on a nonseating surface and neither valve. body had to be-replaced because of corrosion.

Also, affected portions of.the waste gas system were replaced where necessary.

The remainder of the reactor coolant system and interfacing systems which i

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.were inspected, within the limitations of the inspection method employed,

' disclosed no defects attributable to sulfur-induced corrosion.

Therefore, based on the above, the staff finds that GDC 1, 14, 15'and 31 have been met, and that reasonable assurance exists that the public health and safety is protected."

3.3 OTSG Examinations to Determine Extent of' Deoradation In the SER, we summa'rized in Table 3.3-1 an extended post-repair eddy current

' inspection plan and imposed a license condition for monitoring and plant shutdown if primary to secondary leakage increased significantly. By Topical Report 008, Rev. 3 and letter dated September 30, 1983, the licensee indicated that the post-repair baseline eddy current inspection has been completed acco~rding to Table 3.3-1 in the SER. However, the number of tubes tested in each category va'ry slightly, because Table 3.3-1 is intended to provide approximate numbers.

The results of the post repair inspection.were consistent with the 100 percent inspection record in 1982.

. Additionally, the licensee stated that primary to secondary leakage monitoring has been conducted and that a le.akage rate of C.1 gpm above baseline is detectable.

The ir. formation provided does not alter our conclusion that the post-repair extended inserv. ice inspection program is acceptable. However, the license conditions need to be revised to reflect the above information and completion of the OTSG hot functional test program.

The following should be added before the Conclusions in Section 3.3 of the SER:

"The post-repair baseline eddy curent inspection described in Table 3.3-1 was completed both in scope and methodology.

The results of the post-repair

' baseline eddy current inspections were found to be consistent with the 100 percent inspection record of 1982. While some tubes were plugged as

.a preventive measure as a result of the baseline inspection, it is generally

! concluded that small arc length partial through-wall cracks existing in the

~ tubing are not growing nor are new cracks occurring.

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TMI-1 STEAM GENERATOR COMMISSION MEETING e

DECEMBER 7, 1983 9

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TMI-l SG CHRONOLOGY-e LICENSEE PROPOSES 50.-59 APPROACH FOR REPIIR APRIL 30, 1982

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-STAFF IDENTIFIES UNREVIEWED SAFETY QUESTIONS & NEED FOR AMENDMENT FOR OPERATION AUGUST 23, 1982.

e LICENSEE' SAFETY EVALUATION OF RETURN TO-SERVICE SUBMITTED

~ DECEMBER 10, 1982 e

SG AMENDMENT SUBMITTED FEBRUARY 2, 1983 e

SG AMENDMENT RESUBMITTE5 (UNDER SHOLLY)

MAY 9,.1983 e

STAFF PREPOSED DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION

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MAY 31; 1983 ISSUED e

C6MMENT P'ERIOD ENDED JUNE 30, 1983 e

ISSUED SAFETY EVALUATION WITHOUT FINAL SIGNIFICANT HAZARDS CONSIDERATION AucuST 25, 1983 f

DETERMINATION-e RE-OPENED COMMENT PERIOD AususT 25, 1983 i

e COMMENT PERIOD ENDED SEPTEMBER 26, 1983 o

FINAL NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION NOVEMBER 18, 1983 4%

177FA ONCE-THROUGH 2

STEAM GENERATOR (OTSG)

LONGITLDINAL VIEW

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UPPER TUBESEET I

INTERNAL AUXILIAR.Y -

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N0ZZLES (32)

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3 FIGURE 15 TMI-1 Steam Generator Typical Cracks l'NCONEL TUBE

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gY WELD CLAD DVERLAY j#""%

TYPICAL CRACKS V(b Y

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,/ STEEL TU BESHEET CRACK CHARACTERISTICS: CIRCUMFERENTIAL BELOW FILET WELD NOT FULL ARC GENERALLY VERY TIGHT PRIMARY SIDE INITIATED

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EXPERIENCE WITH KINETIC / EXPLOSIVE

- EXPANSION PROCESSES (AS AN ALTERNATIVE TO ROLLING) i i

e UTILIZED 'IN. FABRICATION OF ALL COMMERCI AL NUCLEAR STEAM GENERATORS BY. ONE MANUFACTURER TO ELIMINATE TUBESHEET CREVICES (STARTED 1969)

O IMPLEMENTED USE OF EXPLOSIVE WELDED PLUGS IN NUCLEAR STEAM GENERATORS BY EARLY 1970'S

- EXPLOSIVE CHARGE >10 TIMES STRONGER THAN FOR IMl-1 KINETIC REPAIR TECHNIQUE

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e EXPOSIVE WELDING OF PLUGS APPROVED IN ASME BOILER

'AND PRESSURE VESSEL CODE SECTION XI, ARTICLE IWB-4420 4

VENDOR HAD KINETICALLY EXPANDED MORE THAN 3x106 TUBES SINCE 1966 e

APPROVED FOR FABRICATION OF CLINCH RIVER INTERMEDIATE

. HEAT EXCHANGER BY REACTOR DEVELOPMENT TECHNOLOGY (RDT)

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STAND'ARD 9

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s DISPOSITION OF-TUBES IN TMI-1 OTSG'S TOTAL TUBES (AS FABRICATED) 31,062 KINETIC EXPANSION REPAIRED 29,858

- ID INDICATIONS 40%

(17)

- OTHER INDICATIONS (66)

PLUGGED 1,204

- PRIOR TO PEPAIR (347)

DURING REPAIR (857)

EXPANDED 22-INCH (763) l L

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6 TMI-1 STEAM GENERATOR 1.

STRESS CORROSION CRACKING (SCC) RESULTED FROM A COMBINATION

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e CORRODENT (FROM SODIUM THIOSULFATE LEAKING FROM CONTAINMENT SPRAY ADDITIVE TANK) e Su.SCEPTIBLE MATERI AL (INCONEL) e HIGH. STRESS 2.

TEST PROGRAM e

VERIFIED ECT BY DESTRUCTIVE EXAMINATION OF TUBES e

HYDROSTATIC AND NITROGEN LEAK TESTS OF REPAIRED SG e

HOT FUNCTIONAL TEST PROGRAM, INCLUDING 3 RAPID C00LDOWNS, VERIFIED THAT:

1.

NO CRACKS EXIST WHICH ARE LARGE ENOUGH TO

. PROPAGATE 2.

TOTAL LEAKAGE IS 1 GPH 3.

SULFATE' CONCENTRATI0Hs (0.~1 PPri e

15 YEAR LIFE CYCLE FROGRAM COMPLETED e

CORROSION LEAD TESTS - 18 MONTHS AHEAD OF PLANT e

ADDITIONAL VERIFIC'ATION

- FUTURE PLANT HOT FUNCTIONAL TEST, GRADUAL POWER ESCALATION PROGRAM, AND MID-CYCLE ECT D

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7 IMI-1 STEAM GENERATOR PRINCIPALSTAFFSERCONCLdSIONS s

CAUSE AND EXTENT OF CORROSION IDENTIFIED e

STRESS CORROSION CRACKING (SCC) NOT EXPECTED TO RECUR e

ALL DEFECTS HAVE BEEN IDENTIFIED AND REPAIRED e

KINETIC EXPANSION PROCESS EFFECTIVE AND RELIABLE e

TEST AND POST-REPAIR INSPECTION PROGRAMS ADEQUATE, CONFIRM REPAIRS AND CONCLUSIONS e

GDC 1, 14, 15, 31 MET a

STEAM GENERATORS, RCS, AND WGS WITHIN ORIGINAL LICENSING BASIS 4

-e ACCIDENT ANALYSIS RESULTS NOT SIGNIFICANTLY AFFECTED e

REASONABLE ASSURANCE THAT PUBLIC HEALTH AND SAFETY WILL'NOT BE ENDANGERED l

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TECHNICAL SPECIFICATI0N CHANGE

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NFFDED FOR 0PERATION WITH

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REPAIRFD STEAM GFNERATORS

- - - EXISTING I.S. 4.19.4B

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B.

THE STEAM GENERATOR SHALL BE DETERMINED OPERABLE AFTER COMPLETING THE CORRESPONDING ACTIONS (PLUGGING INCLUDING ALL TUBES EXCEEDING THE. PLUGGING LIMIT AND ALL TUBES CONTAINING THROUGHWALL CRACKS)- REQUIRED BY TABLE 4.19.2 AMENDED I.S. 4.19.4B s.

B.

THE STEAM GENERATOR SHALL BE DETERMINED L

OPERABLE AFTER COMPLETING THE CORRESPONDING ACTIONS (REMOVAL FROM SERVICE BY PLUGGING, OR REPAIR BY THE KINETIC EXPANSION PROCESS, 0F ALL TUBES EXCEEDING THE REPAIR LIMIT AND ALL TUBES CONTAINING THROUGHWALL CRACKS) REQUIRED BY-IABLE 4.19.2 O

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STANDARDS FOR-N0 SIGNIFICANT HAZARDS CONSIDERATION 50.92 -ISSUANCE.0F AMENDMENT.

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(B)

IHE' COMMISSION WILL BE PARTICULARLY SENSITIVE '03 A LICENSE AMENDMENT REQUEST THAT INVOLVES IRREVERSIBLE CONSEQUENCES (SUCH AS ONE THAT, FOR EXAMPLE, PERMITS A SIGNIFICANT INCREASE IN

-THE AMOUNT OF* EFFLUENTS OR RADIATION EMITTED BY A NUCLEAR POWER

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~ PLANT).

(C)' IHE COMMISSION MAY MAKE A FINAL DETERMINATION, PURSUANT.

'03 THE PROCEDURES IN 950.91, THAT A PROPOSED AMENDMENT T.0 AN OPERATING LICENSE FOR A FACILITY LICENSED UNDER 550.21(B) OR 550.22 OR FOR A TESTING FACILITY INVOLVES NO SIGNIFICANT HAZARDS CONSIDERATIONS, IF OPERATION OF THE. FACILITY IN ACCORDANCE WITH L

-THE PROPOSED AMENDMENT WOULD NOT:

l (17 INVOLVE A SIGNIFICANT INCREASE IN THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED) OR (2)

CREATE THE POSSIBILITY OF A NEW OR DIFFERENT KIND OF ACCIDENT FROM ANY ACCIDENT PREVIOUSLY EVALUATED,.OR (3)

INVOLVE A SIGNIFICANT REDUCTION IN A MARGIN OF SAFETY.

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SUMMARY

OF DETERMINATION

-I.

OPERATION OF THE FACILITY IN ACCORDANCE WITH THE PROPOSED AMENDMENT WOULD NOT INVOLVE A SIGNIFICANT INCREASE IN THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED.

e COMPONENTS ARE NOT SIGNIFICANTLY MORE LIKELY TO FAIL THAN BEFORE e

COMPENSATORY MEASURES REDUCE PROBABILITY OF ACCIDENT e

RESPONSE TO PREVIOUSLY EVALUATED ACCIDENTS WOULD NOT BE SIGNIFICANTLY ALTERED.

II.

0PERATION OF THE FACILITY IN ACCORDANCE WITH THE PROPOSED AMENDMENT WOULD NOT CREATE THE POSSIBILITY OF A NEW OR DIFFERENT KIND OF ACCIDENT FROM ANY ACCIDENT PREVIOUSLY EVALUATED.

e COMPONENTS NOT EXPECTED TO FAIL SO AS TO CAUSE A NEW OR DIFFERENT KIND OF ACCIDENT e

NO NEW ACCIDENTS IDENTIFIED WHICH COULD BE CAUSED EY OPERATION WITH PLUGGED' TUBES e

NO MECHANISM IDENTIFIED WHICH WOULD MAKE MULTIPLE TUBE FAILURE MORE LIKELY III.

OPERATION OF THE FACILITY IN ACCORDANCE WITH THE PROPOSED AMENDMENT WOULD NOT ' INVOLVE A SIGNIFICANT REDUCTION IN A MARGIN'0F SAFETY.

8 MARGINS OF SAFETY IN SYSTEMS AND COMPONENTS AND IN PLANT RESPONSE TO ACCIDENTS HAVE NOT BEEN SIGNIFICANTLY REDUCED e

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