ML20012B827

Application for Tech Spec Change Request 199 to License DPR-50,modifying Unscheduled Steam Generator Tube Insp Requirements After Primary to Secondary Leak in Excess of Limits of Spec 3.1.6.3

ML20012B827
Person / Time
Site:	Three Mile Island
Issue date:	03/12/1990
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20012B825	List: ML20012B824 ML20012B827 ML20012B843
References
NUDOCS 9003160384
Download: ML20012B827 (9)

Text

, .;

.)

..y.,. .j "e,

y

+

. . l d

g , ME1FOPOLITAN EDISON CEMPANY JERSEY CEN1RAL POWER & LIGNP CIMPANY .

1 i

PENNSYLVANIA ELECIRIC CIMPANY j i

'IHREE MILE ISLAND NUCLEAR STATION,.INIT 1  !

, l

[:

r Operatirq License No. DPR-50 Docket No. 50-289 +

L 'Ibchnical Specification Charge Request No.199 .I l

t

. ' Ihis Technical Specification' Change Request is submitted in support of Licensee's -

i request to change Appendix 'A to Operating License No. DPR-50 for '1hree~ Mile Island L ' Nuclear Station, thit 1. . As a part of this request, sq-:*1 replacement pages

• l for Appendix A are also included.

1 ~,

\

l GPU NUCLEAR 00RFORATION j l

Ey:

Vice President. & Director, 'IMI-1 Sworn and subscribed th .

to before me'this /d .

day of %4xlu , 1990.  ;

'~'

Y1>

Notary Public ,

NotsialSeal

-Unda L Fuer, Public 7 Mddletown Boro, County. '

. MyCommission Epires 26,1994 Memew,PennsytvansAmof Natanos jq O D

P

t e ,

- Enclosure 1 V_ .

I e .

I. TECHNIchL SPECIFICATION GANGE RIDUEST (TSCR) NO.199 GFW requests that the followirg charged replacenant pages be inserted into the existing 'm&nical Specifications:

Revised pages: 4-78, 4-79, 4-80, 4-81 and 4-82

%ese pages are attamed to this chanje request.

II. REhSW KR WANGE

'Ihis dange is requested to modify the M-1 Technical Specifications for uns&eduled steam generator tube inspection requirements after a primary-to-secondary leak in excess of the limits of Specification 3.1.6.3.- The gt,--=-1 &ange specifies that: (1) when a leakirs tube is located in Group A-1 (" lane wedge" area) all tubes in this group in only the affected steam generator need be inspected (current 'Nmnical Specifications are not explicit in this regard) to inoltu$a those portions of the bihan where the leak was found, and if the results of the inspection fall into the C-3 Category, additional inspections will' be performed-in the same group in the other steam generator; and-(2) when the leaking tube is not in Group A-1, an inspection will be performed on the affected steam generator in armniance with Table 4.19-2.

III. SAFE 1Y EVAT.IATION JUSTIFYING GANGE

'IMI-1 Te&nical Specification Section 4.19.3.c.1 currently specifies that additional uns&eduled inservion inspections shall be performed on each steam generator in armrdance with Table 4.1S-2 during_ shutdown following a primary-tcHieuadary tube' leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification.3.1.6.3.

'Ihe st,- =-i change to limit the' unscheduled inservice inspection to the leaking steam generator following primary-to-iiruadary leakage through the steam generator hihan which avemarlarl Technical Specification limits will both rach= personnel radiation aw-_%

anarciated with the inspections which is consistent with AIARA goals, and provides adequate an=2rance of steam generator tube integrity.

'Ihis provision does not reduce the effectiveness of the overall unscheduled steam generator tube inspection sv:nmu. If the leaking tube is located in the "3ane wedge" area and the results of the unscheduled inspection of the affected steam generator fall into the' c-3 category,4 additional inspections will be performed in the same tube group in the other steam generator. If the leaking tube is not located in the " lane we&J e" area the unscheduled inspection will be performed on the affected steam generator only, in armrdance with existing Technical Specification Table 4.19-2.

I w ,_

^ --

e e

Enclosure 1  !

f ,

page 2 of 4 B

orSG irukastry experience has shown that the " lane wedge" area has been b experiencing corrosion, fatigue, and frutting wear. This area is more susceptible to damage due to the prucimity to the open lane which

' allows hi@er moisture carryover and highest cross flow since the steam

&anges direction from vertical to horiscutal to exit the steam generators.. Performing Te&nical Specification limited tube inspection in the area'where leaks are found enhances plant safety.by identifying

' potential additional tubes whi & may be experiencing similar wear and enabling appropriate vu.a.ctive action to be taken to prevent further 3

tube leakage.

- This ensue =& is consistent with the orSG experience, irdustry experience as endorsed by EPRI in the PWR Inspection Guidelines, and is ,

similar to a request previously approved by the NRC for the Ooonee 1, 2 [

ard 3 plants.

IV. H2.

GPUN~has determined that the Te&nical Specifications Change Request ,

involves no significant hazards considerations as defined by NRC in .

10CFR50.92. [

1. Operation of the facility in accordance with the sur-ed amendment would not involve a significant increase in the probability of occurrence or the consequences of an accident previously evaluated. T The p -fd amendment limits the unscheduled.insarvice inspection "

to the leaking steam generator following primary-to-secondary leakage through the steam generator Hihan which exceeded Te&nical ,

Specification limits. The p --W amendment also limits this unsctwhiled inspection to the " lane wedge" area when the leaking tube is located in this area. The design basis accidents related to this change are widents related to steam generator tube .

integrity. The probability of nmirrence or the consequences of a steam generator tube rupture accident or a main' steam line break accident, which assumes a 1 gpm primary-tc s ilary leak rate, are not incraaaai since adequate assurance of steam generator tube integrity is maintained by the s -f ==d change. - Limiting the -

unsatwhiled inservice inspection to the affected steam generator has '

no adverse affect on the mia?>acy of steam generator. tube integrity. Limiting the unscheduled inservice inspection to the tubes in the " lane wedge" area when the leaking tube is in this area enhances plant safety by identifying potential additional hiham u which may be experiencing similar wear, ocerosion, or fatigue.

Anavs iate acrrective actions are taken to prevent further dopadation. The p - W &ange has no effect cm the inspection methods or acceptance criteria; nor does it reduce the effectiveness of the overall unscheduled steam generator tube inspection rawaru.

Therefore, this change does not increase the probability of occurrence or the consequences of an accident previously evaluated.

> $ [ , ,

Enclosure 1 page 3 of 4 a

- 2. Operation of the facility in ammdance with the puri.ed amendment .'

would not create the possibility of a new or different kind of-accident from any accident previously evaluated. 'Ibe rur s.ed seendment limits the ur./$arhiled inservice-inspection to the leaking steam generator following primary-tc henry. leakage through the steam generatx tubes whidt has amaariari Te&nical' Specification limits. 'Iha p ----9 amendment also limits the unedwhiled inspection to the " lane wedge"' area when the leaking tube is located

- in this area. 'the proposed &ange has no affect on the inspection methods, nor does it rarna the effectiveness of the overall unedwhiled steam generatx tube inspection program. 'Ihe p ,-: :1 changes are related to steam generator tube integrity and tube rupture accidents only, whidt have been analyzed previously.

'Iberwfore, the dange has no effect on the possibility of creating a new or different kind of accident fram any accident previously l evaluated.

3. Operation of the facility in accordance with the r --:=1 amardment '

would not involve a significant reduction in the margin of safety. '

'Iha p =i amendment limits the unediarhiled inservice inspection ,

to the leaking steam generatz following primary-to-secondary "

leakage through the' steam generator tubes whidt a=aiari Tedinical l Specification limits. h p , =i amendment also limits the unsdwhiled inservice inspectica to the " lane wedge" area when the leaking tube is located in this area. Adequate ammirance of steam generat m tube integrity is maintained and plant safety is enhanced l by identifying potential additional ~ tubes whidt may be experiencing similar wear, . corrosion, or fatigue in the area whidt is suscoptible to sudt degradation. Aan.uriate vuu. ?dve actions are taken to prevent further degradation. Performing a 100% inspection of the

" lane wedge" area tubes following a tube leak in excess of the

'DPdinical Specification limits enhances plant' safety by identifying-l; tubes with similar degradation. Ibe prcposal has no effect on the

inspection methods or acceptance criteria, nor does it rarn a the L effectiveness of the overall unscheduled steam generator tube inspection p up.mu. 'Iherefore, it is concluded that operation of '

the facility in aoocrdance with the p urised amendment does not L; involve a significant reduction in a margin.of safety.

h M== ion has provided guidelines pertaining to the application of these standards by listing specific emmples in 457R14870. 'Ihe L p ,-:"i amendment is considered to be the same category as emmple (i) of amendments that are~ considered not likely to involve significant hazards consideration in that the prq = ad change constitutes an.

administrative dange to Technical Specifications. Limiting the unedwhiled inservice inspection to the leaking steam generata is 4 considered an administrative change since it provides clarification that the inspection is required only for the leaking steam generator.

Limiting the unscheduled inservice inspection to 100% of the " lane-wedge" area when the leaking tube is located in this area is considered an. administrative L.

r I

.. _ - -- .~ . . . - -

d

. . y

. Enclosure 1 4 . page 4 of 4 ,

. J l

t dange since it fmmaa the appropriate inspection to the area that has i experienced the most vulnerability to tube damage. 'Ihus, operation of i the facility in aoocrdance _with the p-- =i amendmerit involves no significant hazards'ocnsiderations.

V. IMPulMENRTICN It is requested that the amendment authorizing this &arge hamna effective upon imanunce. In ceder to prevent unrwmamary tube inspections during any future forced steen generator tube leak outage due to the lack of clarity of the requirements of the specification, the request for an interim waiver of ocupliance to Te&nical Specificaticri Section 4.19.3.c. has been provided as outlined in the ,

cover-letter to this amendment cfiange request, to be in dfect until i issuance of the p ,- e-$ amendment. As requested above this TSCR should be p -"3 on an exigent basis.

,s l

l-1 l

l

_________m_

. .. l 1 N

' Enclosure 2 l Waiver of Ctepliance j of 7MI-1 OISG Tabe Ineervice Inspection Recmb---nt frun khich Relief Is Rar=ianted Relief is requested frca the requirement to perfom an additional unscheduled iAmervice inspection of 6% of the tubes in the affected steam generator following a ehutdown for primary-tc s imry " lane wedge" tube leaks'in excess of the limits of specification 3.1.6.3 (1 gpn total for both steam generators).

Technical Specification Change Request 199 (TSCR) proposes that if the leaking tube is identified to be within the boundary of the " lane we&pe" area (tubes in rows 73 to 75 and 77 to 79 adjacent to the open inspection

~

i lane oogrise the " lane" and those in the areas bourded by tubes 66-1, 75-1 and 75-15 and tubes 77-1, 77-15 ard 86-1 ocaprisirq the " wedge"), . in lieu L of meetirq the g soit Twinical Specification inspection requirement (i.e. .i TS 4.19.3.c.1) all tubes in the " lane wedge" area will be inspected.

Event Cirr-tarv'an and Naad For Prr-t NRC Action on March 6 at 0912, 3MI-1 began a plant shutdown because of a primary.to secondary leak in the A OrSG. This m'wred shortly after a' refueling outage. The plant was shutdown and was subcritical at 10:42. It was cooled down at a rate to minimize possible inct= aam in tube leakage.

Following cooldown,-the A OrSG was opened and'a bubble test performed on March 8th. The test identified . tube 1 in row 77 (designated A77-1) as the leaking tube. This tube is in the ." lane wedge" region of the OFSG and had been B3dy Current examined in January 1990 as part of the BR refueling irrarvice inspection gu:pmu. The BR inspection' identified no recordable  ;

' indication of degradation on tube A77-1. Post-leak R$dy Current inspection performed on March 9 identified that'A77-1 had a through wall' defect at the point where the tube exits the botta of the upper tube sheet. .

The failure of tube A77-1 at 7MI-1 on 3/6/90 was identified as a i circumferential1y oriented apptracimately 360* crack. This is believed to l- be the result of envim.-Eally assisted high cycle fatigue (HCF). This l' belief is Maari on ECT data and visual examination of tube A77-1 and on a ocuparison of the A77-1 failure and prior industry experience with ,

envis -Lally assisted HCF crackiry of OrSG tubes. Refer to the table  ;

provided in Attachment 1. The detemination that " lane wedge" area t,iMa

,i L

are ananaptible to this failure mechanism is Maarl on four tube sanples frcan Oconee Nuclear Station removed and analyzed durirq the period from 1976 to 1982.

The failure was unforeseen and unavoidable because R$dy Current testing (ECI) performed for the inservice inspection su:gcun on tube A77-1 during January 1990 yielded no recordable indications of degradation. B&W industry experience confirms that this type of failure emws rapidly ard therefore

_ _ = _ _ _ = _ - - _ _ - _ _ _ - _ - - _ - - _ - _ _ _ _ _ - - _ _ _ _ - - - _ - _ _ . _ _ . - - ,

0 i' 1 e o.

by . 1 Enclosure 2 'l

~ '

page 2 of 4  ;

u Iv 1-evidence of the condition may exist only shortly before leakage would be L experienced.- Tube inspection te &niques do not effectively identify HCF l

precurear coniitions unless they are performed just prior (e.g. hours) before tube failure. Mitigating actions in response to tube leakage are- ,

provided by Plant Normal and Bnergency Prma4nis. l GPLN has ocupleted an inspection of all tubes in the " lane wedge" area of the A OPSG in the area where the A77-1 leak was found. No imperfections of these tubes except A77-1, were identified whi& differ frun the 8R Outage inspection results. It should be noted that ais' defective tube (A78-28) was found during examination of the " lane wedge" area follwing the A77-1 tube leak. The defect was a shalls inside diamater pit " called" at 41% .

through wall hamad on a less than 1 volt one coil indication on a 8x1

• l absolute ECT probe. A review of the 8R ECT data confirms that this defect l existed then and was not called due to its very lw signal level and ,

L shall s phase angle.

l GPLM considers the A77-1 tube failure to be caused by HCF, an irdustry lL

identified problem as characterized above. Since additional eddy current inspection in the " lane wedge" area of the A OPSG has essentially ll. duplicated the results'of inspections performed during the 8R outage h inservice inspection, it is te&nically un-ry to expand the present ECT beyard the " lane wedge" and perform an inservice inspection as required by our existing Technical Specifications section 4.19.1.a.

L Prmpt NRC action on this matter is required to prevent delay of the

(- restart of the unit whi& would otherwise be rsanaaaary to perform testing h per existing Te&nical Specifications. Additional testire adds negligible l safety benefit and provides no significant additional information to minimize'the possibility of recurrence of this event.

C-iEEatory Actions Taken bv GRE to Assure OISG Intearity ,

L

(' Ib asumire the integrity of the A OPSG, GPLM has performed drip ard bubble E

tests to identify all possible leaking tubes.and coupleted addy current testing of all tubes in the " lane wedge" area between the upper tube sheet and the 14th support plate. Except as noted above, all test and inspection results were satisfactory. Tubes A77-1 and A78-28 were plugged.

Safety Sianificance of the Walver of n = 11ance The requested Waiver of Ocspliance permits resJuption of plant operation

. Without completing all Te&nical Specification required ECT examinations of l;

the affected OTSG.- Rather, a fmmad ECT program has been perfanned which inspects all unplugged " lane WeckJ e" area hihaa, which OPSG industry

, experience has demonstrated are prone to the HCF failure mechanism. 'Ihis ECT has been performed down to the 14th support plate which includes all of the HCF failure-prone tube portions. This inspection resulted in the indications noted above. A drip test of the entire OTSG was performed to provide additional confidence in the integrity of the hihaa, and sh med no problems. A post-repair bubble test will also be performed.

t.

~ - -

Enclosure 2 page 3 of 4 1 l

1 Industry experience indicates that the failure of tubes in the area of tube A77-1 are due to HCP. . 'Ihis me&anism is a rapid failure me&anism and precursors may exist only k iefly before failure coours. 'Ihere is no method to predict failures with su& a rapid' development. T a=W=

monitoring is an effective and safe means of detection and operator actier, i a &ieves safe plant conditions, aapeating random ECT of the A OPSG outside ;I the " lane wedge" area would provide no additional technical information j relevant to the current failure me&anism. i i

i e

Duration of the Waiver of h14mnoe (WDC)

It is requested that this WDC remain in effect until such time as the NRC approves the attached TSm. Further, we request that the TSCR be p.(-wi

i. on an exigency basis to support the plant operational schedule. We anticipate your approval of the TSG within about 25 days of the date of.

this letter.

No Sianificant Fawds ConsMarations (NSHC) for the Waiver of h 11ance GEW has concluded that no significant hazards considerations are created by this WDC in that:

l'. 'Ihe WOC does not involve a significant increase in the probability or consequences of OPSG tube rupture. All te&nically relevant ECT has u been performed to inentify tubes whi& may be affected by the HCF '

L me&anism. 'Ihus the probability and consequences of OPSG tube' failure L

are not incraaaM by waiving ECT in non-relevant regions of the OPSG.

2. 'Ibe WOC does not create the possibility of a new or different kind of '

accident since OrSG tube ruptures have been evaluated and the WOC does

! not significantly change their nature, probability, or cu sequences.

^

L, 3. 'Ihe WOC does not involve a reduction in the margin of safety since it in no way reduces the required structural skeyth of the OPSG tubes or the reactor coolant systen.

Envirsc== ital Cansiderationg .

'Ibe NOC has no inpact en envirs===hl consideration in that:

L 1. No changes in effluent limits or types are involved,  !

2. No changes in accident dose a asequences or releases are involved,

3. No routine effluent releases are changed in granting of the WOC, and

4. No increase in power level is involved.

Finally, the Plant Review Group has reviewed this WOC pursuant to TS 6.5 and has concluded that it is acceptable.

.1 j

, . 1 Enclosure 2 I

+

page 4 of 4 L High' Cycle Fatigue Experience  !

, Ctapariscm Table ,

f ,

.c 11 D Industry Experience 'DE-1 Experience 1.. Prahnhantly a ." lane wedge" Tube A77-1 is a periphery tube inned -

tube phencunanon in OPSGs. intely adjacent to the open tube lane.

, t I

2. Crac4cs at the socordary face of B$dy current examination of A77-1 the upper tube sheet and/or 15th reports a through wall, circumferen-(upper most) tube support plate tially oriented crack at the h 4ary L D were so.udary side initiated, face of the upper txte sheet on all 8 .

i oriented circumferentially and of ooils of an 8x1 probe. Seconday side >

L 2 4 coils circumferential extent initiation cannot be established frun (when measured by 8x1 ECT probe). the BCP data.

L 3.. On at least twelve prior ocoa- A77-1 was inspected on January 31, (

sions, N haa which previously had 1990 and found free of recordable l been tested by ECT ard found free indication of shgadations at ,

-of defect indications at the Urs the Urs secondary face following aawidary face have ="WMntly both standard differential and 8x1 L

leaked in service following resunp- absolute ECT exams. Re-review of the 4 tion of plant operation. January 1990. data with.the knowledge of the failure location confirms this area to have no' recordable indicaticms of degradation, p t i 4. Isaks in the lane wedge tubes A77-1 tube leak occurred within 3 days '

due to envirss==d. ally assisted HCF of the startup from the 8R Refuelirg-have often nmwed within 30 days outage. 'Ihe short time to failure after startup frcan a refueling or frun a known defect free condition maintenance outage (on 7 nmaaicms is consistent with the HCF failure in $10 days after startup). mechanism. 'Ihis is the first 'IMI-1 tube to fail via HCF.

t

-- - -- ,. ,. w - - --a e e-----