ML20140C481

Discusses Exams of Core Shroud Welds & Tie Rod Repair Assemblies Conducted During Current Refueling Outage at Plant,Unit 1 & Initiation of Adjustments to One of Core Tie Rods

ML20140C481
Person / Time
Site:	Nine Mile Point
Issue date:	04/11/1997
From:	Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:	Diaz N, Dicus G, Shirley Ann Jackson, Mcgaffigan E, Rogers K, The Chairman NRC COMMISSION (OCM)
Shared Package
ML20140C487	List: ML20140C481 ML20140G898 ML20140G929
References
NUDOCS 9704150150
Download: ML20140C481 (12)

Text

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• • • • • April 11, 1997 I MEMORANDUM T0: Chairman Jackson Commissioner Rogers l  :

Commissioner Dicus '

Comissioner McGaffigan Commissioner Diaz , ,

FROM

L. Joseph Callan Executive Director f Operations 1

SUBJECT:

INFORMATION ON CORE SHROUD CRACKING During the current refueling outage at Nine Mile Point Unit 1 (NMP1), the licensee conducted scheduled examinations of the core shroud welds and tie rod repair assemblies, and initiated adjustments to one of the core shroud tie rods. The tie rods were installed at the last outage as a preemptive repair to potential core shroud circumferential welri cracking, and adjustments to ,

one tie rod were necessary to correct a prev' .as misalignment problem. During '

this inspection, the licensee discovered cracking in several vertical core >

shroud welds. as well as degradation of several components of the tie rod assemblies.

.The NRC staff has been in frequent contact with the licensee and the industry-  ;

sponsored BWR Vessel Internals Project (BWRVIP) respectively regarding the  ;

specific findings at NMP1 and the potential generic aspects of the NMP1 '

findings. The staff has performed a preliminary safety assessment of the i act of vertical weld cracking, and at this time, does not believe it is a si nificant safety issue that warrants any imed10te action by the staff.

With regard to the observed degradation associated with the NMP1 tie rod ~

assemblies. NMP1 has not progressed far enough into its root cause evaluation to determine the reason for the observed tie rod degradation. Until this i

evaluation is completed, the staff cannot determine if the observed '

degradation is unique to the NMP1 design or generically a)plicable. However, it should be noted that the tie rod assembly design at NM)1 is different from other BWRs. due to the unique design of the shroud sup> ort associated with the non-jetpum)BWR-2 design. Also, two other plants wit 1 tie rod repairs (Hatch Unit 1 and >11 grim) were inspected during the Spring 1997 outage and no degradation was found in the repairs. Finally the staff's preliminary evaluation of potential consequences associated with the observed degradation also concludes that there is no significant safety problem warranting  :

1 mediate staff action.

Contact:

C. E. Carpenter. NRR/DE (301) 415-2169 1

(W s 7

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.- The Commissioners -

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i Because of potential interest in this issue by the public and others, my staff j has prepared background information on the issue of core shroud cracking, both 5 general and specific to NMPl. in the form of questions and answers (Attachment). This information may be useful to you and others in responding to inquires on this matter.

l

Attachment:

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i- Boilina Water Reactor (BWR) Vessels Internals Questions and Answers' j

Background

1

{ lihat is the core shroud and what does it do?

. f The core shroud in a BWR is a stainless steel, cylindrical component within l the reactor pressure vessel (RPV) that surrounds the reactor core. The  ;

core shroud separates feedwater in the reactor vessel's downcomer annulus i i region from the cooling water flowing up through the reactor core. In l addition, the core shroud laterally supports the fuel assemblies to l l maintain control rod insertion geometry during operational transients and ,

4 postulated accidents. For GE BWR-3 and later designs (i.e., BWRs with jet pumps), the core shroud also provides a refloodable volume for safe 1

shutdown and cooling of the reactor core during postulated accident t conditions. In BWR-2 reactors (e.g., Nine Mile Point 1 [NMPl] and Oyster J. Creek), the core shroud is vertically supported by a truncated conical core

! shroud support ring, which is welded to the core shroud at one end and to the reactor vessel wall at the other. In the BWR-2 design, the core would be cooled by spray cooling during a postulated design basis event.

4 (Figures 1 and 2 are diagrams of a typical core shroud.-)

How are core shrouds constructed?

5 i

BWR core shrouds are typically constructed from three shroud shells (the upper, middle and lower shrouds shells), and two support ring structures (the top guide support ring and core support rings). Some designs, such as j .the core shroud design at Pilgrim, have.an additional support ring i structure. The core shroud shells are typically fabricated from welded, type 304 or 304L stainless steel plates. The ring supports are. fabricated

. from either plates or ring forgings, of type 304 or 304L stainless steel.

i Fabrication of BWR core shrouds involves both axial and circumferential 4

welds to form shells and ring segments. The structural configuration of

! core shrouds in GE BWR-2 designs is similar to later designs, with the

exception that the shroud is supported by a truncated conical support ring.

[ What are other BWR internals constructed off

.Many internal components of a BWR vessel are made of materials similar to i the core shroud. A few examples are austenitic stainless steels, alloy j 600, alloy X750, and alloy 182 weld metal.

lihat is intergranular stress corrosion cracking (IGSCC)?  !

, IGSCC is a time dependent material degradation process, which is known to

be caust

d, and accelerated, by the presence of crevices, residual stresses, i material ' sensitization, irradiation, cold work, temperature, and corrosive environments. The heat affected zones (HAZ) of welds are particularity

{

susceptible to IGSCC. In a memorandum dated January 4, 1994, the NRC staff reported to the Commission that IGSCC of the internal components of BWRs j- was emerging as a technical issue. The core shroud was one of the internal i components listed in the memorandum as being susceptible to IGSCC. IGSCC l has also been detected at welds in other BWR components, including core  !

Attachment I t

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L-l l5 spray spargers, feedwater spargers, jet pump hold-down beams, top guides, core support plates, and access hole covers. Nuclear licensees have r implemented inspection and repair programs to ensure continued structural integrity of these components.

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l When was IGSCC of BWR internals first reported?

1~

In 1990, GE reported the occurrence of cracking in the core shroud of a 4 foreign BWR. The shroud cracks were located in the HAZ of a-l circumferential. core shroud weld in the reactor's beltline region. The

! reactor had completed approximately 190 months of power operation before 1 the flaw indications were discovered. The first cracking at a U.S. plant i was found by the Carolina Power and Light Company (CP&L) at Brunswick i Unit 1 (BSEP1) in early July 1993. The most extensive flaw indication 11 i the BSEPl shroud was located on the inside shroud surface of the H3 weld in

! the HAZ. The examinations also revealed circumferential cracking along j significant portions of welds H1 and H2 (using conservative assumption, up l

to 74% and 68% of the weld circumferences, respectively). In addition, j CP&L reported minor cracking associated with the HAZs of circumferential welds H4, H5, H6a, and H6b. Analyses of the cracks at the H1, H2, and H3 welds indicated that structural margins would still be maintained for the l next operating cycle. Nonetheless, CP&L opted to modify the core shroud in i order to ensure the structural integrity of the H2 and H3 welds during f normal operating, transient and postulated accidental loading conditions.

The modification. involved installing a series of GE-designed mechanical

clamps around the H2 and H3 welds to provide an alternative load bearing t' capability in lieu of the H2 and H3 welds. The NRC accepted the j design for implementation on January 14, 1994.

l Out of 32 BWR licensees, 27 have inspected their shroud welds per GL 94-03 ,

(see below) and 5 installed preemptive repairs without inspecting (8 other '

{ licensees installed a repair after inspecting). Seventeen licensees found  !

i some degree of cracking in their shrouds; however, no cracking reported to j date has exceeded ASME Code allowable. To date, no licensee has found a 360' through-wall crack of any shroud circumferential weld.

. What is the safety implication of core shroud degradation?

L

! The NRC staff evaluated potential safety concerns associated with the L possibility of a 360' circumferential separation of the shroud following a l postulated loss-of-coolant accident (LOCA). The NRC staff considered the i potential for separation of.the shroud during postulated accidents.

Postulated separation could either prevent full insertion of the control i rods, or open a gap large enough to preclude the emergency core cooling systems from fulfilling their intended safety functions. The accident

! scenarios of primry concern were the main steam line break and the f recirculation line break. The more serious event associated with j postulated 360* through-wall cracks in the upper shroud circumferential i

welds (e.g., H2, H3) is the steam line break, since the lifting forces ,

f generated may be sufficient to elevate the upper portion of the shroud and

the top guide and potentially cause difficulties with rod insertion. The J

recirculation line break-is the more serious event associated with cracks in the lower elevations of the core shroud. The recirculation line break f Attachment 1 i~

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is a greater concern at lower weld elevations because this type of LOCA has the potential to result in a lateral displacement of the shroud. Such a lateral displacement.of the shroud could affect the ability of control room .

operators to insert control rods into the core and could prevent adequate (

core cooling by not allowing the core te reflood following a LOCA.

In consideration of the consequences of a 360* through-wall failure of the shroud coincident with a LOCA, the NRC staff has r.onservatively estimated the risk contribution from shroud cracking and cetermined that it does not pose-a high degree of risk at this time. However, the NRC staff has also determined that structural margins specified in the American' Society of Mechanical Engineers Boiler and Pressure Vessel (ASME BPV) '. ode could be exceeded if the cracks were sufficiently deep and continued propagating through the shroud during normal operating, transient or accident conditions, possibly resulting in the loss of a layer of the defersc-tn-depth strategy. Therefore, the NRC staff concluded that it was appropriate for BWR licensees to implement timely inspections and/or repairs of their core shronds. To implement this position, the NRC staff issued Generic Letter (GL) 94-03, dated July 25, 1994, requesting BWR licensees to inspect their core shrouds by the next refueling outage and justify continued safe operation until inspections can be completed. This position enabled the NRC staff to verify compliance with the inservice inspection requirements of Section 50.55a of Title 10 of the Code of federa7 Regulations'(10 CFR 55.55a), and ensured that the risk associated with core shroud cracking

. remains low.

What regulatory actions have been taken to address core shroud cracking?

~

On July E5,1994, the NRC issued GL 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," to all BWR licensees (with the exception of Big Rock Point, which does not have a core shroud).

Yhn GL requested licensees to:

1. In:pect cere shrouds no later than the plant's next refueling outage.

2. Perform materials related and plant-specific safety analyses with respect to the core shrouds. -

3. Develop core shroud inspection plans, which address inspection of all core shroud welds and take into account the latest available technology developed by the industry for inspection of BWR internal components.

4. Develop plans for core shroud evaluation and/or repair.

GL 94-03 further required BWR licensees to submit to the NRC staff a schedule for their first core shroud inspection, a safety analysis supporting continued operation of the facility until inspections were conducted, one or more drawings of the core shroud configurations, and a history of core shroud inspections comaleted to date of GL 94-03. The NRC staff also requested that licensees summit, no later than 3 months before inspecting or repairing of their core shrouds, the scope of the planned core shroud inspections and their plans for evaluating and/or' repairing Attachment I

-- . .- -- - ~ - . --

4 their core shrouds based on inspection results, and the core shroud inspection results within 30 days of completing the shroud examinations.

1 The responses to GL 94-03 were reviewed and plant-specific safety I evaluations were written for each BWR, as well as a generic safety '

i evaluation for the BWRVIP's core shroud repair design criteria. While the various plant-specific and generic safety evaluations concentrated on the effects of circumferential weld cracking, it was understood that "...

repairs shall ensure the displacement of any weld that experiences 360*

through-wall cracking under. normal, upset, and faulted loading conditions 4

does not exceed the values established to: 1) ensure bypass leakage is limited .. and, (2) limit the deflection an(d deformation of interr,01s to

, ensure ... that design control rod drive scram capability and [ECCS) functions are not affected." These evaluations are summarized in NUREG-1544, " Status Report: Intergranular Stress Corrosion Cracking of BWR Core l

Shrouds and Other Internal Components," dated March 1996, attached. The report also provides a summary of activities to the date of publication, the NRC staff's conclusions, and future hetions planned. The NRC staff is preparing a Supplement to NUREG-lE44, to be published by Fall 1997.

What actions have the industry taken to resolve BWR internals concerns?

The BWR Owners Group (BWROG) submitted its criteria for evaluating BWR core shrouds in a letter to the NRC dated April 5,1994. The inspection strategy detailed in the BWROG report focused on a ranking system that bases a plant's IGSCC susceptibility according to its age, construction l materials, and reactor coolant conductivity level. The BWROG concluded.  !

that it was unlikely that any development of cracking would fail 'to satisfy  ;

the safety margins specified in Section XI of the ASME Code; however, both l the BWROG and ?ndividual licensees have indicated that preemptive repairs '

would be implemented in highly susceptible plants to assure that ASME Code margint continue to be met.

In 1994, the BWROG formed a new industry organization, the Boiling Water l

Reactor Vessel and Internals Project (BWRVIP) to address the issue of age-  ;

related degradation of BWR internal components. This organization is  !

designed to ensure that the BWRVIP's efforts are reviewed on both the  !

technical a6d executive levels, and to encourage widespread industry acceptance of BWRVIP guidelines, criteria, and methods. To date, the BWRVIP has submitted for NRC staff review 14 guideline reports dealing with inspection, evaluation and repair of various BWR internal components and systems; an additional 24 reports are planned to be submitted to the NRC for review and approval over the next year.

- What are current regulatory activities?

Priorities have been assigned for the development and review of the industry topical reports. The intent is that when the NRC staff reviews of the BWRVIP reports are complete, that these reports will form an acceptable basis for licensee inspections of BWR internals. Adoption of the guidelines in these reports will either be volunta'ry by licensees or 4

possibly through a generic letter. In the interim, the NRC staff is monitoring licensee activities on a plant-specific basis.

Attachment 1

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j, How is the core shroud repair modification (tie red assembly) designed? ,

f-j The core shrou'd repair hardware consists of a set of tie rod assemblies which are also referred to as stabilizer assemblies. Each stabilizer assembly consists of an upper spring, an upper support bracket, a tie rod, j a mid-span tie rod support, t lower spring, a lower anchor assembly, and  !

other minor parts. The tie rods provide the vertical load carrying l'

, capability from the upper bracket to the lower anchor assembly attached to i the shroud support plate at the bottom of the shroud. They keep the shroud j clamped together, assuming t5at the horizontal welds in the shroud are i fully cracked. Most tie rod' installations have been designed by GE and ,

j consist of four stabilizer assemblies located 90' apart. MPR Associates  !

l has also designed a tie rod repair. The MPR tie rod design for the  ;

i FitzPatrick and Oyster Creek core shrouds consists of 10 tie rod assemblies "

that function in a manner similar to the GE design. At Vermont Yankee, the  !

! MPR designed installation has four tie rod assemblies similar to the GE  :

L arrangement. j

(

In a typical design the vertical locations of the radial springs were chossn to provide the maximum support for the shroud, top guide, core I plate, and, the fuel assemblies. The upper spring provides radial load carrying capability from the shroud, at the top guide elevation, to the RPV. The lower spring provides radial load carrying capability from the shroud, at the core support, plate elevation, to the RPV. The upper stabilizer bracket provides an attachment feature to the top of the shroud as,well as restraint of the upper shroud welds. The mid-span tie rod support is installed to provide a limit stop for the central shroud cylinder if the top and bottom circumferential welds on the cylinder are

. assumed to have cracked 360* through-wall. At the top, each stabilizer assembly fits through slots which are machined into the shroud head. The stabilizer assembly supportu the upper spring and has a hole through which the tie rod passes. The tie rod is held against-the upper bracket with a nut. The tie rod extends downward along the length of the shroud and is threaded into the lower spring. The lower spring has a pin at the bottom, which is. attached to the clevis in the lower support. The lower support is anchored near the bottom o? the core shroud in several different arrangements. At NMP1 the lower support is bolted to the shroud support cone with two toggle bolts. The primary forces that the stabilizers would

' experience are from seismic events, LOCA differential pressure loads, and differential thermal expansion.  ;

The stabilizers are installed with a small tensile pre-load to ensure that all components are tight. The stabilizer assemblies are further thermally I pre-loaded during normal operating conditions. This tensile thermal  :

preload in the tie rod results from the thermal expansion coefficient for l the new M abilizer hardware being less than the thermal expansion l coefficiant of the shroud. The upper and lower springs of the stabilizers j are installed with a small radial pre-load such that they provide radial '

support for the shroud. During normal operation, the shroud and stabilizer i springs radially expand due to thermal growth slightly more than the RPV, l which increases the radial pre-load and assures that the springs provide lateral support for the shroud during normal operation.

l Attachment 1 l

i l

i -

i -

Nine Nile Point 1 Plant Snecific i j What was found at Nine Mile Point 1 (NMPI)? ,

i Significant cracking was discovered in the vertical welds of the core l shroud at NMP1 after one cycle's operation following installation of the <

j preemptive core shroud repair in lieu of a cor+rehensive examination of the  :

shroud. Further, the clamp devices  !

also degraded after one cycle's opera tion, (tieincluding rods) used for theclip a failed repair andwere an  :

, out-of-position lateral wedge support on one tie rod, discolored clips on l two other tie rods, and a loose nut on the fourth tie rod.

i

+

l .Why was the licensee repairing one of the tie rod assemblies?  !

Because the tie rod was mispositioned during the initial installation, the  :

'. lower lateral support of this tie rod was partially resting on the inside '

radial surface of the recirculation outlet nozzle instead of the vessel I

' wall. The NRC staff reviewed the licensee's evaluation of the mispositioning and found operation of NMPI in this condition acceptable for  !

one operating cycle. The licensee was conducting pre-planned inspections of the tie rods and core shroud welds as well as operations to repair the i mispositioned tie rod when it discovered the degradation of the tie rods.

t l What is the status of the licensee's inspection activities at NMPl?

i

! Niagara Mohawk, the licensee for NMPI is presently performing examinations l

. of the core shroud and is evaluating the results. The licensee has  !

informed the NRC staff that it expects to complete its evaluations during )

, the week of March 31, 1997, and that it will submit the results to the NRC 1 l staff for review and approval prior to NMP1 restart.

What actions'will the NRC take prior to allowing NMP1 to restart?

The NRC staff is closely following the activities at NMP1 and will evaluate  !

the licensee's actions prior to allowing restart from this outage. The NRC l staff will review the licensee's root cause determination regarding the 1 observed degradation of the core shroud and the tie rods and will provide a written safety evaluation prior to restart.

Considering that the NMP1 shroud contained vertical and circumferential cracking and that portions of the tie rods assemblies were observed to be degraded, would the shroud have been able to perform its design safety functions?

With regard to the performance of the stabilizer assembly, it appears that the loose nut would most likely have caused a reduction to the clamping -

force on the shroud at that location. However, the remaining three tie rods would have provided adequate thermal preload during normal operation.

There would have been a negligible increase in the postulated crack opening and leakage during a postulated steam line break accident scenario.

The failure of the retainer clip is not likely to render the tie rod repair completely ineffective. The tie rods continue to provide vertical clamping Attachment I l

. forces to compensate for potential circumferential cracking. HIwever,

, lateral support from the' bumper on the degraded tie rod would be lost at the location where the spring clip failed and the lateral wedge support dislodged. This could reduce the seismic margin for the shroud; however, lateral support is provided along several other locations. In addition,-

i the core support plate wedges restrain core plate movement during a seismic i event, thereby assuring control rods insertion, even if the horizontal

! welds are cracked through-wall.

From the results of the examinations of the shroud welds performed to date, it a pears that sufficient ligaments exist in both the circumferential and axia directions to allow the NMP1 core shroud to perform its intended safety functions even if the tie rod assemblies would have completely failed. However, considering the degradation of the tie rods, the licensee needs to complete its inspections and evaluate the shroud structural

' integrity to ensure that these conclusions remain valid.

l Are the stresses from the tie rod assembly gonsidered a driving force for the i cracks in the vertical welds and a factor in accelerating crack growth at l#lPl?

! Stresses would be induced on the vertical shroud welds as a result of clamping of the shroud by the tie rods. The licensee has stated that preliminary analyses show that the clamping stresses on the vertical welds during normal operation are too low to be a factor in influencing crack growth. The licensee will evaluate other factors, such as fabrication stresses and weld residual stresses. The clamping stresses from the repair assembly do not appear to be a concern in this regard.

What caused the failure of the lower wedge support retainer clip in one tie rod assembly at NMPl?

The licensee has not yet determined the root cause. Several failure scenarios are being investigated, including a potential installation deficiency.

Attachment 1 I

i .' j I Generic considerations Considering.the above, why shouldn't all BWR's be immediately inspected for (

this degradation? ,

1 Most of the IGSCC susceptible BWR's have inspected their core shroud I j circumferential welds at least once between the Fall of 1993 and the Spring l l- of 1996. If the integrity of the circumferential welds is assured, then a *

, crack that might occur in, or adjacent to, a vertical weld could extend l l

axially from one.circumferential weld to the next without affecting the

[

structural integrity of the shroud. Further, analysis shows that a crack [

l of this magnitude would not open sufficiently to cause leakage of water  ;

during a design-basis event-that could compromise core re-flood capability. .

[ Of the thirteen plants that have installed preemptive repairs, all have

conducted, or will have conducted within their current refueling outage,  ;

sufficient inspections /reinspections of shroud welds, either vertical or j, circumferential, to assure that they maintain their structural integrity.  :

What has the NRC staff and BWRVIP done regarding inspection activities at

} other BWRs based on the findings to date at NMPl?

J i

[ The NRC staff has held discussions w!th the BWRVIP via a telephone i

conference call on March 20, 1997. During the conference call, the BWRVIP j stated that it has recommended to its u mber utilities that are either t i presently in or are scheduled to shortly enter a refueling outage to  !

j consider, in light of the results from NMP1, to expand its inspection scope

- of.the core shroud to concentrate on the high fluence areas around the i shroud belt line, and to include examinations from the shroud outside l diameter (OD). The BWRVIP further stated that they are evaluating a ,

revision to its " Guidelines for Reinspection of BWR Core Shrouds (BWRVIP- l

! 07)" document, which is presently under NRC staff review. The NRC staff '

{ continues onsite inspection of the NMP1 activities and is interacting i j frequently with the licensee and representatives of the BWRVIP. The NRC  !

! staff is publishing an Information Notice to inform licensees of the-

! cracking of the vertical ~ welds at NMPl. Also, the NRR Projects staff has i i contacted the BWR licensees presently in, or about to enter, refueling >

l outages to inform these licensees of what has been found to date at NMP1 and suggest that the licensees review the information for applicability to their facilities and consider actions, as appropriate, to detect or avoid siellar problems.

Based on the known deficiencies identified in the repair assembly, what would

[ the safety consequences have been if a design basis accident event occurred?

The consequences from a safety standpoint appear to be minor. In all cases

, where a preemptive repair has been installed to date, licensees'

inspections have indicated that sufficient ligaments in the shroud

! circumferential welds exist to ensure integrity in the near term. Since

. -the shroud circumferential welds continue to maintain shroud integrity,

) cracking of the axial welds in any ring or shell segment will not j compromise overall shroud integrity nor will cause a leak sufficient to

! prevent core reflood capacity.

l

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NUREG-1544 Status Report: Intergranular Stress Corrosion Cracking of BWR Core Shrouds and Other Internal Components U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation t,, ~'

AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications l

Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room, 2120 L Street, NW., Lower Level, Washington, DC 20555-0001

2. The Supenntendent of Documents, U.S. Government Printing Office, P. O. Box 37082, Washington, DC 20402-9328

3. The National Technical information Service, Springfield, VA 22161-0002 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.

Referencec occuments available for inspection and copying for a fee from the NRC Public Document Roo~ nclude NRC correspondence and internal NRC memoranda: NRC bulletins, circulars, ento mation notices, inspection and investigation notices; licensee event reports; vendor repo*ts and correspondence: Commission papers; and applicant and licensee docu- i ments anc correspondance. '

The f ollomng occuments in the NUREG series are available for purchase from the Government Pnnting Of t.ce formal NRC staff and contractor reports, NRC-sponsored conference pro-ceedings, enternational agreement reports, grantee reports, and NRC booklets and bro- )

Chures. Also available are regulatory guides NRC regulations in the Code of Federal Regula- l tions, and tiuciear Regulatory Commission Issuances. )

Documents avaitab4e from the National Technical Information Service include NUREG-series  ;

reports anc technical reports prepared by other Federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

1 Documents avaitable from public and special technical libraries include all open literature '

items, such as books, journal articles, and transactions. Federal Register notices, Federal l and State legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertations, foreign reports and translations, and non-NRC con- I ference proceedings are available for purchase from the organization sponsoring the publica-tion cited.

Single copies of NRC draft reports are availt.ble free, to the extent of supply, upon written I request to the Office of Administration, Distribution and Mail Services Section, U.S. Nuclear I Regulatory Commission, Washington DC 20555-0001. I l

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, Two White Flint North 11545 Rockville Pike, Rock-ville, MD 20852-2738, for use by the public. Codes and standards are usually copyrighted )

and may be purchased from the originating organization or, if they are American National i Standards, from the American National Standards Institute,1430 Broadway, New York, NY 10018-3308.

n

NUREG-1544 d

Status Report: Intergranular Stress Corrosion Cracking of BWR

Core Shrouds and Other Internal l

Components 4

i a

I i

1 Manuscript Completed: March 1996

Date Published

March 1996 I

i

J. Medoff, NRC Technical Monitor Division of Engineering

Oflice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 l

4

,p" %,

(M..... ),,

2

ABSTRACT On July 25, 1994, the U.S. Nuclear Regulatory This report summarizes the staff's basis for issuing Commission (NRC) issued Generic Letter (GL) 9443 to GL 94-03, as well as the staff's assessment of plant-obtam infonnstion needed to assess compliance with specific responses to GL 94-03. The staff is continually regulatory requirements regarding the structural integrity evaluating the licensee inspection programs and the of core shrouds in domestic boiling water reactors results from examinations of BWR core shrouds and (BWRs). other internal components. This report is representative of submittals to and evaluations by the staff as of This report begins with a brief description of the safety September 30,1995. An update of this report will be significance of intergranular stress corrosion cracking issued at a later date.

(IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the U.S. and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components.

iii NUREG-1544

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t r.

l CONTENTS -

Page Abstract......................................................................................................................... iii ListofFigures.................................................................................................................. ix ListofTables................................................................................................................... xi List o f A ppendices . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x ii i  !

4 l

ExecutiveSummary........................................................................................................... xv i l

i Acknowledgements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xv ii 1 introduction................................................................................................................. 1-1 1 2 BWR and Core Shroud Desi gns . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 BWR and Core Shroud Design Characterstics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1

! 3 i

t 2.2 Construction Materials and Fabrication Methods .............. . .. .. ... ..... . . . .......... .. .. ....... .. .. . . .. . 2-1 i

3 Intergranular Stress Corrosion Cracking of BWR Internal Components ........................................ 3-1 i l

l 4 BWR Core Shroud Cracking - Systems Evaluation and Safety Assessment ................................... 41 l 4.1 Structural Integrity Assessments . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 l

4.2 Safety Significance of 360' Through-Wall Cracks During Normal Operations

! and Operational Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.3 Safety Significance of 360' Through-Wall Cracks During Design Basis Accidents...................... 4-2 5 BWR Core Shroud Cracking - Summary of Significant Operating Experience ............................... 5-1 i

! 5.1 Crackmg at a Foreign BWR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 l 5.2 Crackmg at Bmnswick Steam Electric Plant, Unit 1 ... ... .... .. ...... ........ .... .. ... . ................ ...... 5-1 5- 1 5.3 Crackmg at Commonwealth Edison Plants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

l 5.4 Crackmg at Oyster Creek Nuclear Generation Station ....................................................... 5-2 5.5 Crackmg at Vermont Yankee Nuclear Power Station ........... ................... ................. ...... 5-2 i

6- 1 i 6 ladustry Efforts to Address the IG SCC lasue . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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i 6.1 Genenc Approach Taken to Address the IGSCC Issue ....................... .... ... ............ ... ... ....... 6-1 i 6.2 Efforts by the Boiling Water Reactor Owners Group ............. .......................................... 6-1 3

i' v NUREG-1544 f

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Page 6.3 Establishment of the Boiling Water Reactor Vessel and Internals Project ................................ 6-1 6.4 Activities of the General Electric Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-2 6.5 Activities of the Electric Power Research Institute .. .......... ............. ....... ............. ......... .... . 6-3 7 GL 9443, 'Intergranular Stress Corrosson Cracking of Core Shrouds in Boiling Water Reactors" ....... 7-1 7.I Content of GL 94-03...............................................................................................7-1 7.2 Generic Assessment of the Industry's Responses to GL 94-03.............................................7-1 8 Plant-Specific Assessments and Results of Core Shroud Inspections or Repairs ............................... 8-1 8.0 Overview.............................................................................................................8-1 8.1 Boston Edison Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.1.1 Assessment of the Response to GL 94-03 for the Pilgrim Nuclear Power Station .............. 8-1 8.1.2 Repai r of the Pilgrim Core Shroud . . ... .. . . . .. . .. . ... . .. . . . . . . . . .. . . . . . . . . . . .. . . . . . ... . . . . ..... ... . . . . .. 8-3 I 8.2 Carolina Power and Light Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-3 8.2.1 Assessment of the Response to GL 94-03 for Brunswick Unit i ......................... ......... 8-3 1

8.2.2 Reinspection of the Brunswick Unit 1 Core Shroud .................................................. 8-3 8.2.3 Assessment of the Response to GL 94-03 for Brunswick Unit 2 .................................. 8-3 8.3 C&-wealth Edison Category "C" Plants . . ....... ... .. . . ... . .. . . . . . . . . .. . .. . . . .. . . . . .. ... . . . . . ..... .. . .. . .. 8-4 8.3.1 Assessment of the Response to GL 94-03 for Dresdon Unit 3 and Quad Cities Unit 1........ 8-4 I

8.3.2 Assessment of the Response to GL 94-03 for Dresden U iit 2 and Quad Cities Unit 2......., 8-4 8.3.3 Repairs of the Dresden, Units 2 and 3, and Quad Cities, Units I and 2, Core Shrouds ...... 8-5 8.4 General Public U tilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 -5 8.4.1 Assessment of the Response to GL 94-03 for the Oyster Creek Nuclear Generation Station 8-5 )

u 8.4.2 Inspections and Repair of the Oyster Creek Core Shroud ........................................... 8-6

'l 8.5 Georgia Power Company . . . . . . . .. . . . . . . .. . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 I 8.5.1 Assessment of the Response to GL 9443 for Edwin 1. Hatch Unit 1............................ 8-6 1 I

!' 8.5.2 Repair of the Edwin 1. Hatch Unit 1 Core Shroud .... . ... ..... .... . ...... ... . ................ ....... 8-6 NUREG-1544 vi 1

1

Page 8.5.3 Amaaammant of the Response to GL 94-03 for Edwin I. Hatch Unit 2 ...... ..................... 8-6 8.5.4 Repair of the Edwin 1. Hatch Unit 2 Core Shroud .............,, .................. ................. 8-7

8. 6 I ES Utilities, Inc. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-7 t

8.6.1 Ammaammant of the Response to GL 94-03 for the Duane Arnold Energy Center .............. . 8-7 3.6.2 laspection of the Duane Arnold Core Shroud .................................... .................... 8-8  ;

8.7 ) !ebraska Public Power District . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-8 8.7.1 Assesament of the Response to GL 94-03 for the Cooper Nuclear Station ................ ...... 8-8 8.7.2 laspection Scope for the Cooper Core Shroud . ...... . . ................... ... . .... ...... ............. 8-8 8.8 Niagara Mohawk Power Corporation Category 'C' Plants .................................................. 8-9 8.8.1 Am=a== ment of the Response to GL 94-03 for Nine Mile Point Unit 1............................ 8-9 i 8.8.2 Repair of the Nine Mile Point Unit 1 Core Shroud ................................ .................. '8-9 4

8.9 Northeast Nuclear Energy Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-9 ~!

8.9.1 Assessment of the Response to GL 94-03 for the Millstone Unit 1 Core Shroud ............... 8-9 ,

1 1

8.9.2 Reinspection Scope for the Millstone Unit I Core Shroud ......................................... 8-10  :

8.10 Northern States Power Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 10 1

)

8.10.1 Amaaaamant of the Response to GL 94-03 for the Monticello Nuclear Generation Station ... 8-10 B.10.2 Inspection of the Monticello Core Shroud ..................... ..... .... ... ........... ................. 8-10 8.11 Philadelphia Electric Company Category 'C' Plants ........................................................ 8-11 l i

8.11.1 A=ma== ment of the Response to GL 94-03 for Peach Bottom Atomic Power Station Unit 2 8-11 1

8.11.2 laspection of the Peach Bottom Unit 2 Core Shroud ................................................ 8-11 8.11.3 Amaaaament of the Response to GL 94-03 for Peach Bottom Atomic Power Station Unit 3 8-12

8. I1.4 .einspection Scope for the Peach Bottom Unit 3 Core Shroud ...................... ............ 8-13 8.11.5 reach Bottom Core Shroud Repair Designs ........................................................... 8-13 8.12 Power . Authority of the State of New York . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . 8 13 8.12.1 Aasessment of the Response to GL 94-03 for the James A. FitzPatrick Nuclear Power Plant 8-13 vii NUREG-1544

Page 8.12.2 Repair of the James A. FitzPatrick Core Shroud .................................................... 8-14 8.13 Tennessee Valley Authority . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 - 14  ;

1 3.13.1 Assessment of the Response to GL 9443 for Browns Ferry Nuclear Units 1,2, and 3 ..... 8-14 8.13.2 Inspection of the Browns Ferry Unit 1, 2, and 3 Core Shrouds .................................. 8-14 8.14 Vermont Yankee Nuclear Power Corporation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 15 8.14.1 Assessment of the Response to GL 94-03 for the Vermont Yankee Nuclear Power Station 8-15 ,

8.15 Plants with Category " B " Core Shrouds .... . . . .. . . .... . . . . . .. . . . . . . . . . . . .. . . . .. . . . . . . . . .. . . . ... .. . .. . . . . . . . .. . 8- 15 8.16 Plants with Category ' A ' Core Shrouds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8- 16 9 IG SCC in Other BWR Internal Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9- 1 ,

9.1 Core Plate and Top G uide Cracking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9- 1 9.2 Jet Pump Hold-Down Beams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 -2

9. 3 Access H ole Covers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 10 Conclusions and Future Actions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. . . . . . . . . . . . . . . . .. . 10- 1 1

1 11 References.................................................................................................................Il-1 l L

NUREG-1544 viii

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1 LIST OF FIGURES l

, 1 l l page '

l Figure 2.1-1 Reactor Vessel Flow Paths in GE BWR-3, BWR-4, BWR-5, and BWR-6 Reactor Designs 2-3 l

l Figure 2.2-1 Structural Configuration Typical of GE BWR-3, BWR-4, i BwR 5, and BwR-6 Core Sa,oud Desi. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-4 l

Figure 2.2-2 Core Shroud Weld Locations Typical of GE BWR-3, BWR-4, BWR-5, and B WR-6 Core Shroud Designs. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-5 l Figure 6.3-1 Typical Modification Design Proposed for Repair of BWR Core Shrouds ...... ................. 6-5 l

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i ix NUREG-1544 i

LIST OF TABLES Page Table 6.2-1 BWRVIP Susceptibility Rankings and Core Shroud Inspection Recommendations ......... .. .. 6-4 xi NUREG-1544

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i APPENDICES  !

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Page i l

1 1~

A List of BWROG Members and BWRVIP Subcommittees .... ........... ...... .. . .. ... .. .... . .. .. ............ .... A-1 B Plant-Specific BWR Core Shroud Summaries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 1 C List of BWR U tilities and Reactors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C- 1 D Abbreviations and Nomenclature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D 1 E List of Staff Contributors to N U REG- 1544 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-1 i 1

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xiii NUREG-1544

EXECUTIVE

SUMMARY

Many internal components of boiling water reactor intended safety functions. The accident scenarios of (BWR) vessels are inade of materials susceptible to primary concern are the main steam line break and the intergranular stress corrosion crackmg (IGSCC), recirculation line break. b more serious event including stamiess steel, alloy 600, alloy X750, and associated with cracks in the upper shroud welds (e.g.,

alloy 182 weld metal. IGSCC is a time dependent H2, H3) is the steam line break, since the lifting forces material degradation process, which is known to be generated may be sufficient to elevate the top guide and caused and accelerated by the presence of crevices, potentially cause difficulties with rod insertion. The residual stresses, material sensitization, irradiation, cold recirculation line break is the more serious event work, and corrosive environments. As operating BWRs associated with cracks in the lower elevations of the core age, the numhar of cracking incidents is expected to shroud. 'Ihe recirculation line break is a greater concern increase. - The U.S. Nuclear Regulatory Commission at lower weld elevations because this type of LOCA has (NRC) staff has been meeting every year since 1988 the potential to result in a lateral displacement of the with the Boiling Water Reactor Owners Group and the shroud. Such a lateral displacement of the shroud could General Electric Company, and later with the Boiling affect the ability of control room operators to insert Water Reactor Vessel and Internals Project, to review control rods into the core and could prevent adequate the generic safety implications of reactor internal core cooling.

components that are considered to be susceptible to IGSCC. In consideration of the consequences of a 360* through-wall failure of the shroud coincident with a LOCA, the In 1990, crack;ng of the core shroud was visually NRC has conservatively estimated the risk contribution observed in an overseas BWR. The cracking was from shroud cracking and determined that it does not lemaad in the heat-effected zone of a circumferential pose a high degree of risk at this time. However, the weld in the beltline elevation of the shroud. Cracking of NRC has also determined that structural margins BWR core shrouds reported at U.S. plants in 1993, specified in the American Society of Mechamical 1994, and 1995 has boca the most significant cracking of Engineers Boiler and Pressure Vessel Code could be BWR internal components. h core shroud is a exceeded if the cracks were sufficiently deep and stainless, steel cylinder that separates the feedwater in the continued propagating through the shroud during normal reactor vessel's downcomer annulus region from cooliag operating, transient or accident conditions, possibly water flowing through the reactor core. The cor,e resulting in the loss of a layer of the defense-in-depth shroud also performs the important functions of properly strategy. Therefore, the staff has concluded that it is directing coolant flow through the core and maintaining appropriate for BWR licensees to implement timely the core geometry. For GE BWR-3 and later desig=., inspections and/or repairs of their core shrouds. To the core shroud also provides a structural boundary to implement this position, the NRC staff issued Generic allow for reflooding of the reactor core to two-thirds letter (GL) 94-03 (July 25,1994), requesting that BWR core height under postulated accident conditions. licensees inspect their core shrouds by the next refueling outage and justify continued safe operation until Significant circumferential cracking has been discovered inspections can be completed. This position enabled the at the Brunswick Unit 1, Dresden Unit 3, Quad Cities staff to verify compliance with the inservice inspection Unit 1, Oyster Creek, and Vermont Yankee nuclear requirements of Section 50.55a of Title 10 of the Egds stations. In light of the extent of cracking observed at of Federal Reenlations, and ensured that the risk these plants, the staff evaluated potential safety concerns associated with core shroud cracking remains low, associated with the possibility of a 360* circumferential -

separation of the shroud following a postulated lons-of- As of early September 1994, the NRC staff received all coolant accident (LOCA). The staff considered the of the BWR licensee submittals in response to GL 94-03.

potential foi separation of the shroud during postulated The staff has completed its evaluations of the licensee accidents to either prevent full insertion of the control responses and has transmitted the safety evaluation rods, or open a gap large enough to preclude the reports to the appropriate BWR licensees. The staff emergency core cooling systems from fulfilling their concluded that, for all cases, BWR licensees had xv NUREG-1544

_ provided sufficientjustification to operate their facilities Nine Mile Point Unit 1, and Pilgrim nuclear plants.

until core shroud inspections or repairs could be Repairs will be made at additional plants ifinspection implemented. The staff based its conclusions on the results indicate that large scale cracking of following factors: circumferential shroud welds has occurred, or may be made at the discretion of the licensee in lieu of (1) No 360o through-wall core shroud cracking has comprehensive core shroud examinations (pre-emptive been observed to date in any U.S. BWR at which core shroud modifications). These repairs or the licensee performed a shroud inspection. modifications are designed to ensure the structural ,

integrity of the core shrouds based on an assumption that (2) All analyses performed by U.S. licensees to date the shroud circumferential welds are completely cracked, indicate that, even if cracking did exist in a and am being reviewed by the NRC staff on a case-by-particular BWR core shroud, sufficient ligaments case basis.

would remain in the shroud so that the structural  !

integrity of the shroud would be ensured for the In the spring of 1994, the industry formed a new [

remainder of the plant's operating cycle. organization, the BWR Vessel and Internals Project (BWRVIP), to address the issue of IGSCC of BWR (3) No U.S. BWR has exhibited any of the symptoms internal components. The BWRVIP is headed by several (power-to-flow mismatch) that would be indicative high level utility executives to ensure that top executives t of leakage through a 360o through-wall shroud in the industry are aware of its function, purpose and crack. efforts. The BWRVIP subsequently submitted i documents a&lressing an integrated safety assessment of  ;

(4) Main steam line or recirculation line breaks are the issue, guidelines on performing nondestructive  !

both considered to be low frequency events. examinations (NDE) of core shroud welds, guidelines on inspection scopes for BWR core shrouds, and generic (5) There were only short durations until core shroud guidelines and acceptance criteria in regard to '

inspections were to be conducted or repairs were performing flaw evaluations and repairs of BWR core to be implemented by the individual BWR shrouds. The NRC staff has approved the generic repair licenseen. criteria document, the latest revision to the BWRVIP guidelines regarding core shroud inspection scopes and To date, core shrouds have been repaired (modified) at flaw evaluations, and the BWRVIP guidelines regarding i the Brunswick Units 1 & 2, Hatch Units 1 & 2, core shroud NDE methods.

FitzPatrick, Oyster Creek, Quad Cities Units 2, i

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NUREG-1544 xvi

ACKNOWLEDGEMENTS

'the authors wish to thank the many persons who Reactor Owners Group, the Boiling Water Reactor contributed i====arably to the work documented in Vessel and Internals Project, the Electric Power this report and to the quality of the final product. We Research Institute, and the General Electric Company.

would like to thank Mr. C.E. Carpenter and Mr. D.S. In addition, we would like to thank Ms. J. Beeson of the Brinkman, who were instrumental in coordinating the Technical Editing Staff, and Ms. A.D. Lowery and Ms.

staff activities with members of the Boiling Water J.E. Brooks for proofreading this document.

xvii NUREG-1544

l 1 INTRODUCTION in a memorandum dated January 4,1994, the staff of the his report also includes a number of appendices.

U.S. Nuclear Regulatory Commission (NRC) reported Appendix A lists the industry members of the Pailing to the Commission that intergranular stress corrosion Water Reactor Owners Group (BWROG) and Boiling ces: king (IGSCC) of the internal components of boiling Water Reactor Vessel and laternais Project (BWRVIP) water reactor (BWR) vessels was emerging as a Subcommittees. Appendix B presents plant-specific core technical issue (Ref.1). The core shroud was one of the shroud data sheets summarizing important information f internal components listed in the memorandum as being provided by the licensees to assist the NRC in its susceptible to IGSCC. evaluation of core shroud cracking. Appendix C lists die licca.ees, along with their cosiWing BWR units.

On July 25,1994, the NRC issued Generic Letter (GL) Appendix D presents a list of acronyms, abbreviations 94-03, "latergranular Stress Corrosion Cracking of Core and scientific units used in this report. Appendix E lists Shrouds in Boiling Water Reactors," which requested the staff members who have contributed to the staff's that BWR licensees inspect their core shrouds at the assessment of the issue of IGSCC in BWR internal earliest refueling outage (RFO) for their plants (Ref. 2). components.

Since then, most BWR licensees have mspected or repaired their core shrouds during planned RFOs. These This report is representative of submittals to and inspections have shown that core shrouds can crack at evaluations by the staff as of hy='--- 30,1995. An s circumferential weld locations. IGSCC has also been update of this report will be issued at a later date.

detected in other BWR components, including core spray spargers, feedwater spargers, jet pump hold-down beams, top guides, core support plates, and access hole covers. Nuclear licensees have implemented inspection and repair programs to ensure continued structural integrity of these components.

His report presents background information, current status, and future actions needed to address the issue of IGSCC in BWR internal components. Chapter 2 of this report describes BWR core shroud design characteristics and fabrication materials. Chapter 3 discusses the mechanism of IGSCC in BWR internal components.

Chapter 4 presents an assessment of the safety significance of postulated accidents with 360' through-wall cracks. Chapter 5 of this report summarizes significant BWR cracking events in the industry.

Chapters 6 and 7 of this report summarize the industry and NRC efforts taken to date to address the IGSCC issue. Chapter 8 summarizes the staff's assessment of the industry's plant-specific responses to GL 94-03.

Chapter 9 of this report summarizes cracking events that have occurred in other BWR internals to date. Chapter 10 presents general staff conclusions with regard to the issue of IGSCC in BWR internals and addresses future actions to be taken. Chapter 11 provides a list of references.

11 NUREG-1544

2 BWR AND CORE SHROUD DESIGNS 2.1 BWR and Core Shroud Desien Characteristics means of maintaining emergency core cooling during postulated loss-of-coolant accidents (LOCAs), when The core shroud in a BWR is a stainless steel, ECCS actuation is necessary to maintain reactor safety.

cylindrical component within the reactor pressure vessel Recirculation of the emergency coolant back to the (RPV) that surrounds the reactor core. De core shroud annulus region of the vessel, in this case, occurs by way separates feadwater in the reactor vessel's downcomer of the jet pump diffusers. The height of the diffusers annulus region from the cooling water flowing up provides for a two-thirds core height re-flood capability through the reactor core. In addition, the core shroud of the reactor core. These designs allaw for re-flood of laterally supports the fuel assemblies to maintain control the core in a relatively short time. Figure 2.1-1 depicts rod insertion geometry during operational transients and the reactor vessel flow paths typical of GE BWR-3, postulated accidents. For GE BWR-3 and later BWR-4, BWR-5 and BWR-6 reactor designs, and designs, the core shroud also provides a refloodable illustrates how the reactor water level achieved in these volume for safe shutdown and cooling of the reactor designs during normal operating conditions differs from core during postulated accident conditions. that achieved during postulated LOCAs.

The General Electric Company (GE) has been the only in contrast, shon and long term cooling responses of manufacturer of BWRs in the United States. GE models non-jet pump (BWR-2) plants for large recirculation line currently licensed for operation in the U.S. range from breaks rely on core spray, as the vessel will not flood.

BWR-2 reactors with Mark I type containments Recirculation flow enters the reactor yessel from the (drywell-torus type) through BWR-6 reactors with Mark bottom and the ECCS for large breaks are two III containments (drywell-Weir wall type). All GE redundant, double capacity, core sprays. Therefore, the BWR models are equipped with low pressure emergency degree of any resulting cooling deficiency depends on core cooling systems (ECCS), and some form of the final condition of the core spray system. Long term automatic depressurization system. ECCS designs for cooling is unchanged as containment flooding is later BWR models (BWR-4, BWR-5 and BWR-6) also unaffected.

include high pressure spray or coolant injection systems.

2.2 Construction Materials and Fabrication Methods Some distinct differences set GE BWR-2 reactors span from later GE BWR designs (BWR 3 through BWR-6). BWR core shrouds are typically constructed from three in BWR 2 reactors, the core shroud is vertically shroud shells (the upper, middle and lower shrouds supported by a conical core shroud support ring, which shells), and two support ring structures (the top guide is welded to the core shroud at one end and to the support ring and core support rings). Some designs, reactor vessel wall at the other. in later GE BWR such as the core shroud design at the Pilgrim Nuclear models, the core shrouds are supported by core shroud Power Station, have an additional support ring structure.

support legs or cylinden, which are in turn welded to the lower reactor vessel head. Another distinct The core shroud shells are typically fabricated from characteristic of GE BWR-2 reactors is the absence of welded, type 304 or 304L stainless steel plates. The jet pumps and recirculation loops. As a result, the core ring supports are fabricated from either plates or ring in BWR-2 models must be cooled using natural forgings, of type 304 or 304L stainless steel. The circulation, rather than forced recirculation, the method carbon content of shroud plates or forgings fabricated of core cooling used in later designs. from type 304 stainless steel in these shrouds typically ranges from 0.03% to 0.07%. The carbon content of ne absence of jet pumps in BWR-2 reactors also shroud plates or forgings fabricated from 304L stainless precludes the design from having any direct systematic steel is typically less than 0.03 %.

ties between the lower plenum area and the annulus region of the reactor vessel. This is important from a Fabrication of BWR core shrouds involves both axial defense-in-depth standpoint. For BWR-3 and later and circumferential welds. Welding of the core shroud designs, the presence of jet pumps provides an easy shells and ring segments is typically accomplished using 2-1 NUREG-1544 f

1

i shielded metal arc welding (SMAW), automated submerged arc welding (SAW), automated gas tungsten arc welding (GTAW), automated gas metal are welding (GMAW), or a combination of these welding techniques.

Figure 2.2-1 illustrates the structural configuration that is typical of GE BWR-3, BWR-4, BWR-5 and BWR-6 )

core shroud designs. Figure 2.2-2 illustrates the I locations of circumferential welds that are typical of these designs, although the exact number and numerical notation of the shroud welds may vary from plant to plant. The structural configuration of core shrouds in GE BWR-2 designs is similar to later designs, with the exception that the shroud is supported by a truncated conical support plate.

1 NUREG-1544 2-2

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NUREG-1544 2-4

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

l 3 INTERGR'tNULAR STRESS CORROSION CRACKING OF 8WR INTERNAL COMPONENTS Many internal components of a BWR vessel are made of metal during cooling and are tensile in nature. Although i materials that are susceptible to IGSCC, such as stainless weld stresses are not easily quantified, previous steels, alloy 600, alloy X750, and alloy 182 weld metal. investigations indicate that tensile stresses on the weld Core shrouds are among the BWR internal components surface may be as high as the yield stress of the that have been shown to be susceptible to IGSCC. material. The stress decreases to compressive levels in IGSCC is a time-dependent material degradation process the center of the welded section.

which is known to be caused and accelerated by the i presence of corrosive environments, crevices, residual The fabrication process used for the core support and

• I 2 stresses, material sensitization, cold work, and top guide rings can also play an important role in irradiation. IGSCC susceptibility. Current available inspection data indicate that shrouds fabricated with forged ring industry experience has shown that the portions of the segments are more resistant to IGSCC than rings l core shrouds most susceptible to IGSCC are commonly constructed from welded plate sections. The difference l 1 associated with shroud base metal located ia areas in susceptibility relates to differences in the shroud immediately adjacent to the shroud welds. Tinse base fabrication processes. Most plants have support rings metal regions are known as the heat dfe ed r zones fabricated from arc segments that are cut from rolled (H AZa) since they are known to undergo intense thermal plates and welded to form the ring. His process  !

cycling during the welding process. This thermal exposes the short transverse direction in the material to l cycling may cause the HAZs to undergo a phenomenon the reactor coolant. In this case, elongated grains and known as " sensitization". During " sensitization" carbon stringers in the material are exposed to the reactor diffuses to the grain boundaries of the HAZ base metal. coolant environment, thereby increasing the nrnbability This carbon precipitates out at the grain inundaries in for initiation of cracking or cruci !!ke d.

fects. Forged the form of complex chromium carbides upon cooling of rings are typically not cut in this manner, and therefore the weld melt. The precipitation of these carbides do not have the "end grains" exposed to the depletes the chromium in the steel material adjacent to environment.

the grain boundaries. Because the presence and distribution of chromium on the surfaces of the material The degree of reactor coolant water quality also provides corrosion resistance in stainless steels, its correlates strongly with the degree of IGSCC depletion increases the potential for the grain boundaries susceptibility in BWR internal components. Industry become crack initiation sites. experience has shown that shrouds operated in coolants with high ionic conductivities are more likely to be Sensitization of stainless steels typically occurs when the highly susceptible to IGSCC than shrouds operated in steels are exposed to temperatures ranging from 1000*F coolants with low ionic conductivities'. Furthermore, to 1500*F. Temperatures in this range are easily industry experience has shown that reactor coolant chieviole during welding. The degree of sensitization systems (RCSs) operated at highly positive, increases with increasing carbon content of the stainless electrochemical potentials (ECPs) are more susceptible 4 steel materials. By contrast, material resistance to IGSCC can be increased if the carbon content is kept below 0.030%. Therefore, low carbon stainkss steels 1 Conductivity is a measure of the anionic and cationic Effer greater resistance to sensitization, and are therefore content of liquids. As a reference, the conductivity of more resistant to initiation of IGSCC. pure water is ~0.05 pS/cm (~0.05 pmhos/cm).

Reactor coolants with conductivities below 0.20 pS/cm Welding processes can also inc-ease IGSCC (0.20 mhos/cm) are considered to be relatively ion susceptibility by introducing high residual stresses in the free; reactor coolants with conductivities above 0.30 stainless steel material located at the weld joint. Rese S/cm (0.30 mhos/cm) are considerd to have a stresses result from thermal contraction of the weld relatively high ion content.

3-1 NUREG-1544

i to IGSCC than RCSs operated at more negative ECPs8.

The industry has made a considerable effort to improve water chemistry at nuclear facilities over the past 10 years. Industry initiatives have included introducting hydrogen water chemistry as a means of lowering ECPs (i.e., making the ECPs more negative) in the RCS, and introduction and using improved cleanup resins as a ,

means of improving water purity in the RCS. The l effectiveness of hydrogen water chemistry in reducing the susceptibility of core shrouds to IGSCC initiation has l not been fully evaluated; however, its effectiveness in I reducing IGSCC in recirculation piping has been ,

demonstrated.  !

l 2 ECP is a measure of a material's susceptibility to corrosion. in the absence of an externally applied current (as is the case for reactor intemal components in the RCS), the ECP is equal to the open circuit potential of the material. Industry experience has shown that crack growth rates in reactor intemal components are low when the ECP s: ~-0.230 V. j I

l I

l NUREG-1544 3-2

l l

1 4 BWR CORE SHROUD CRACKING - SYSTEMS EVALUATION  !

AND SAFETY ASSESSMENT  !

I l

4.1 Structural Interrity Assessments 4.2 Safety Sienificance of 360* ThrouefcWa))

Cracks Durine Normal Onerstionr and To assess the structural integrity of core shroud welds Operation Transients I with cracks extending up to 360' in the circumferential I I direction, an analyst must consider the effects ofloading In order to provide a bounding consequence assessment conditions, material properties, and crack geometries on for the cracking observed to date, the NRC staff the shroud. The shroud is constructed of stainless steel, postulated complete weld failures at various locations t

which has a high degree of fracture toughness. In fact, during normal power operation. He intent of this the core shrouds were fabricated from 1.5-inch to 2-inch consequence assessment was to demonstrate that fuel I thick plates primarily for stiffness during transport and geometry and core cooling would be maintained given installation. In addition, the operational and postulated the unlikely occurrence of a through-wall failure of any accident loads produce low stress levels in the shroud. horizontal weld, and to identify whether horizontal weld ,

Therefore, as previously described, adequate structural failures would be detectable. In their responses to i 4

integrity for the shroud can he maintained despite Generic Letter 94-03, all licensees expected that any

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extensive err,ckinE. 360' through-wall crack would be detected during i normal operations. l He core shrouds at most U.S. BWRs were not originally designed in accordance with the design rules During normal operations with any horizontal weld of the American Society of Mechancal Engineers sufficiently cracked, some upward displacement of the I (ASME) Boiler and Pressure Vessel Code. However, shroud could occur, depending on the postulated crack the reactor internal compoaents have been included in location, operating conditions, and plant type. A small 1 plant inservice inspectica programs in accordance with amount of lift at the upper shroud weld locations would Section 50.55a of Title 10 of the Code of Federal produce anomalies such as increased coolant

]

Reaulahons (10 CFR 50.55a), and are therefore within temperatures and/or reduced coolant flow. These power the scope of Section XI of the ASME Code. In anomalies, power / flow mismatch, are detectable during ,

addition, the NRC staff and the BWROG are currently normal operations. After detection of such an anomaly,  !

developing a draft Subsection IWG to Section XI of the a normal shutdown is expected to be initiated until the ASME Code, on requirements for reactor vessel cause of the anomaly is determined and corrected.

components and internal structures. Subsection IWG Analogous results have been experienced at other j will augment Section XI by providing additional details operating reactors when the shroud head bolts were l concerning examination, inspection, and acceptance improperly engaged.  ;

standards for flaws in internal components. l During most limiting operating conditions,100% power he approaches in the ASME Code address both the and rated flow, the maximum expected vertical '

linear elastic fracture mechanics of the limiting crack displacement can be postulated based on the pressure (LEFM analysis) and the potential for gross deformation differences across the shroud head and the shroud and subsequent failure of the uncracked material in the support. Shroud lift above the top of the fuel channels vicinity of the crack (limit ic;ad or LLA analysis). To has the potential to affect core geometry and control rod date, in reviewing these analyses for licensing purposes, insertion. In most cases, the maximum postulated the NRC has required that licensees substantiate the use vertical displacement at l'.3 and H4 is not sufficient to ,

of the LLA methodology by examining the stress disengage the top guide from the fuel channels. Shroud intensification present at the crack tip through finite lift at H2 does not affeet core geometry since H2 is element modeling. located above the top guide. Some uncertainty remains l

l l

4-1 NUREG-1544 l

1

n - - - . . - - . _ - . _ - - . . - - - . - - . - - . . - - - _ _ - . _ - - -

i as to whether shroud displacement at the lower welds, postulated 360' through-wall failure of the shroud. The H5 through H8, would be detectable during normal rapid depressurization that is characteristic of this event ,

operations. Postulated shroud displacement, if any, at scenario has the potential to result in loads or moments &

these locations would be small and would not affect the which could cause a lateral displacement or tipping of <

ability to insert the control rods if necessary. the shroud. Such a lateral displacement or tipping of the shroud may affect the ability of plant operators to N consequences of operation during anticipated insert the control rods during the event, and may result operational transients with a 360' through-wall crack in in an opening of the shroud that could allow bypass i the shroud are bounded by the design basis accident leakage through the shroud and out of the pipe break.  ;

analysis in Section 4.3. It is expected that the following Large bypass leakage could potentially affect the ability f

anticipated operational transients could increase shroud of the plant operators to reflood the core, maintain i loads above those experienced during normal operation: adequate core cooling following the pipe break, and shut  !

down the reactor with the standby liquid control system  ;

(1) pressure regulator failure - open; (SLCS).

(2) recirculation flow control failure - increasing to The NRC has raised additional concerns in regard to the f maximum flow; potential for a shroud displacement to damage other  !

vessel internal components during postulated accident .

(3) inadvertent actuation of the Automatic conditions. In particular, a vertical shroud displacement l Depressurization System (ADS). has the potential to damage the core shroud support legs  ;

as a result of the impact loadings that would occur upon !

For a 360o through-wall crack, these loads could lead to resettling of the core shroud. Displacement of a core i complete weld separation and/or result in higher upward shroud also has the potential to damage core spray lines, j displacements than normal operations. All licensee particularly if the core spray lines have been degruded analyses concluded that during such postulated events, prior to the event.

MCPR Safety Limit, lew Water level, and Reactor a Overpressure Limit are not violated. The staff has developed a probabilistic safety assessment [

based on assessments of potential 360' through-wall j 4.3 Safety Sienificance of 360' Through-Wall failures of the circumferential shroud welds in the i Cracks Durine Desien Basis Accidents Dresden Unit 3 and Quad Cities Unit I core shrouds. 1 The assessment estimated the potential contribution to i In' order to anness the significance of potential cracking core damage frequency resulting from the cracked [

beyond that observed to date, the staff has evaluated the shrouds. For the upper shroud welds (e.g., the H2 or!

safety implications of a postulated 360' circumferential H3 welds), the staff concluded that any 360' through- !

separation of the shroud. "Ihe staff has determined that wall failure would be expected to be detected during  ;

the detectability and consequences of 360' through-wall normal operation (e.g., either by power / flow mismatch cracking relate directly to the cracked weld location. or noise monitoring). For lower shroud welds (H5 and ,

The main concern associated with cracks in the upper lower), the staff has concluded that a 360' through-wall j shroud welds arises during a postulated MSLB shroud failure would have to occur concarrent with a concurrent with a 360' through-wall failure of a shroud. large rupture of either a main steam line or a  ;

During such a postulated accident, the resulting recirculation line to be capable of achieving the loading differential pressures are expected to be large enough to magnitudes that could move the shroud. However, it vertically displace the remaining upper shroud assembly. should be noted that probabilistic risk assessments These liRing forces potentially could elevate the top categorize such MSLBs and RLBs to be low frequency guide above the fuel assemblies. The resulting safety events. To date, no MSLB or RLB has ever occurred concerns would include a loss of lateral support for the at an operating nuclear plant, and the unlikely '

fuel assemblies, potential loss of control rod insertion occurrence of a 360' through-wall crack concurrent with capability, and potential damage to the core spray a large pipe break would be necessary to pose any piping, incremental risk. Finally, the shroud may not move in the most adverse manner during these events, and there The main concern associated with cracks in the lower is a good chance that core cooling and reactor shutdown shroud welds arises during a RLB concurrent with a would be achieved with no adverse consequences.

NUREG-154' 4-2

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

l l

l Considenng these =====maats the staff concluded that core shroud cracking does not pose a high degree of risk

' for the short tenu, and that inunediate plant shutdowns were not warranted for inspections. However, the staff concluded that degradation of the core shroud could inspect plant safety if plants with degraded shrouds were

allowed to continue to operate for extended periods.

The staff therefore concluded that 360' cracking of the shroud was a safety concern for the long term based on the following consulerations:

1 (1) .the potential for exceeding the ASME Code structural margins if the cracks are sufficiently deep and continue to propagate through the subsequent operating cycle; (2) the uncertainties associated with the behavior of a 360' through-wall cracked core shroud under accident conditions; (3) . elimination of a layer of defense-in-depth for plant safety.

To date, the majority of BWR industry licensees have conducted inspections, evaluations, and/or repairs of their core shrouds to address the issue of core shroud and BWR internal crackmg. Such . mapection, )

evaluation, and repair activities adequately ensure that j simultaneous failure of multiple internal components will j not result is adverse risk to the general public. I 1herefore, the staff has concluded that the current status of BWR core shrouds and internal components does not constitute a high degree of risk to the general public at j this time. i l

4-3 NUREG-1544 l

I

I i i 5 BWR CORE SHROUD CRACKING -

SUMMARY

l OF SIGNIFICANT OPERATING EXPERIENCE l I

i 5.1 Cructrine at a Foreien BWR of the weld. This circumferential weld joins the top 4

guide support ring to the middle shroud shell. Using i ln 1990, the General Electric Company (GE) reported conservative assumptions, CP&L hypothesized that the i l the occurrence of cracking in the core shroud of a crack could extend nearly 360' around the foreign BWR, b shroud cracks were located in the circumference of the weld. Boat samples taken from the

', HAZ of a circumferential core shroud weld in the H3 weld identified IGSCC as the crackmg mechania, reactor's beltline region. 'Ihe reactor had completed and indicated that the H3 flaws could be more than i approximately 190 months of power operation before the 0.036 m (1.4 in) deep. b VT-1 examinations also 4

flaw indications were discovered. revealed circumferential cracking along significant

,. portions of welds H1 and H2 (using conservative

As a result of this discovery, GE issued Rapid assumption, up to 74 % and 68% of the weld Information Communication Services Information letter circumferences, respectively). In addition, CP&L (RICSIL) 054, " Core Support Shroud Crack reported minor cracking associated with the HAZs of 4 Indications," on October 3,1990 (Ref. 3), to all owners circumferential welds H4, H5, H6a, and H6b.

! of GE-designed BWRs. RICSIL 054 summarized the cracking found in the core shroud of the foreign BWR. GE analyses of the cracks at the H1, H2, and H3 welds 2

It also recommended that nuclear utilities owning BWRs indicated that structural margins would still be with high. carbon steel core shrouds perform a visual maintained for the next operating cycle. Nonetheless,

} exa-ameion of the ace-ible areas of the shroud seam CP&L opted to modify the core shroud in order to 4

(circumferential) welds and associated HAZs on the ensure the structural integrity of the H2 and H3 welds i inside and outside abroud surfaces. during normal operating, transient and postulated i accidental loading conditions. W modification involved I

installing a series of mechanical clamps around the H2 5.2 Crackinn at Bmnswick Steam and H3 welds. These clamps were designed by the

. Electric Plant. Unit 1 General Electric Company (GE) to provide an alternative load bearing capability in lieu of the H2 and H3 welds.

la early July 1993, the Carolina Power and Light The NRC reviewel the clamp design in the Fall of 1993 Company (CP&L) performed enhanced visual testing and accepted the design for implementation on

! January 14,1994 (Ref. 6).-

(VT-1) exannaations of the core shroud at the Brunswick Steam Electric Plant, Unit 1 (BR-1). CP&L conducted j the VT 1 exanunations of the core shroud welds on the j inside and outside surfaces of the shroud, in accordance 5.3 Crackinn at Commonwealth Edison Plants l

with the recommendations of GE RICSIL 054. W l results of the VT-1 examinations revealed the presence Commonwealth Edison Co npany (Comed) reported

) of numerous flaw indications in the core shroud, cracking in the core shrouds of Dresden Unit 3 (DR-3) and Quad Cities Unit 1 (QC-1). Comed discovered )

i GE issued RICSIL 054, Rev.1, " Core Shroud Cracks," these cracks as a result of shroud examinations  !

dated July 21,1993 (Ref, 4), to inform the industry of conducted during the DR-3 and QC-1 Spring 1994 i the cracking in the BR-1 shroud. In addition, the NRC refueling outages (RFOs), h most extensive cracking  !

informed the industry of the BR 1 shroud cracking in at each plant was associated with the H5 weld, wiuch l

l. Information Notice (IN) 93-79, ' Core Shroud Cracking joins the mid-shroud shell to the shroud's core plate l at Beltline Region Welds in Boiling-Water Reactors ' support ring. The licensee's examinations included both I

+

dated Sp kr 30,1993 (Ref 5), enhanced VT-1 and ultrasonic testing (UT) methods.

Using conservative assumptions, Comed deternuned that b most extensive flaw indication in the BR-1 shroud the cracks could extend nearly 360' areund the was located on the inside shroud surface of the H3 weld, circumference of the welds.

CP&L determmed that the crack was located in the HAZ

. 5-1 NUREG-1544 i-t

j To inform the industry of the cracking at DR-3 and refueling outage. GPU's examinations of the OCNGS  ;

QC-1, the NRC issued IN 94-42, " Cracking in the core shroud included UT inspections of accessible areas Lower Region of the Core Shroud in Boiling-Water on shroud welds H1, H2, H4, H5, and H6a, and Reactors," dated June 7,1994, and Supplement 1, dated enhanced visual examinations of welds H3, H6b, and July 19,1994 (Refs. 7 and 8). Instead of opting to H9. The results of the shroud exammations indicated parform an i==antaata repair of the DR-3 and QC-1 core significant cracking at the H4 weld. This weld joins the shrouds, Comed proposed to operate these plants for an upper mid-shroud shell to the lower mid-shroud shell, additional 24 months while they designed and fabricated and is in the vicinity of the reactor beltline region. 'Ihe a per=anant repair. results of the OCNGS shroud examinations also I indicated some minor cracking at the H2 and H3 welds.

To support the conclusion that both of these units could be operated safely, the licensee submitted a safety After completing the UT examinations of the H4 weld, evaluation demonstrating the DR-3 and QC-1 core GPU decided that they would modify the OCNGS core shrouds would meet the following safety criteria: shroud before restarting the unit. GPU submitted its i design for the OCNGS core shroud modification on l (1) 'Ibe existing cracks would not propagate through October 25,1994 (Ref.10). The OCNGS core shroud the shroud wall during the next fuel cycle. modification involved installing a series of tie rod ,

assemblies symmetrically around the shroud. 'Ihese tic ,

(2) h existing uncracked ligaments would continue rod assemblies are designed by MPR to restrict vertical to provide sufficient structural integrity and to and lateral motion of the shroud, assuming that all meet the requirements of the ASME Code. circumferential welds in the core shroud fail coincident i with a design basis event.

(3) The existing cracks would not compromise the safety functma of the shroud under all postulated "Ihe str.ff issued its SER regarding the OCNGS core design-basis accident conditions. shroud modification on Nov. 25,1994 (Ref.1I). 'Ibe I staff concluded that the shroud modification design ,

Comed concluded that the plants could be operated for selected by GPU provides an acceptable alternative load  ;

their full cycles. 'Ihe NRC reviewed the licensee's fisw carrying capability for the OCNGS core shroud. 'Ihe l evaluations and safety ==a=ments regarding the DR 3 staff therefore concluded that the modification design t and QC-1 core shrouds. N NRC also performed was acceptable for implementation. ,

independent analyses of the DR-3 and QC-1 core  ;

shrouds in order to validate the licensee's results and  ;

conclusions. 'Ibe staff based the analyses on a bounding 5.5 Crackine at Vermont Yankee Nuclear Power Station I initial crack depth of 0.033 m (1.3 in) and a bounding l crack growth rate of 3.5E-10 m/s ($E-5 in/hr). W 'Ihe Vermont Yankee Nuclear Power Corporation )

staff did not allow for additional structural margin credit (VYNPC), the licensee for the Vermont Yankee Nuclear created by the inner diameter fillet weld at H5. Upon Power Plant (VY), has completed NDEs of the VY core '

conclusion of the reviews, the staff deternuned that the shroud and feedwater nozzles. 'These examinations were  ;

ASME Code snargins (and thereby, the requirements of scheduled as part of VYNPC's GL 94-03 and '

10 CPR 50.55a) would be satisfied for 15 months of hot NUREG-0619 activities, respectively. Prelimmary operation commencing from the time of the DR 3 and results of VYNPC's core shroud examinations indicated )

QC-1 startups. 'Ibe staff issued its Safety Evaluation that a significant degree of cracking ( = 340' ~ 345* in Report (SER) regardmg the cracking in the DR-3 and circumference) existed in VY core shroud weld H5, 1

. QC-1 core shrouds on July 21,1994 (Ref. 9). which joins the lower mid-shroud shell to the core ,

support ring. VYNPC also determinart that cracking of l a lesser degree was indicated at shroud welds H1, H2, j 5.4 Craciuna at Ovater Cnek H3, H4, and H6. All feedwater nor.zle ena-mations i Nuciant Genernhon Stahon were negative for relevant indications.

General Public Utilities (GPU), the licensee for the VYNPC's flaw evaluatsons of indications in the H1, H2, l Oyster Creek Nuclear Generation Stahon (OCNGS) H3, H4, and H6 welds indicated that the welds have 1 inspected the OCNGS core shroud dunng the Fall 1994 sufficient structural margins for at least two additional NUREG-1544 5-2

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

e operating cycles. In evaluating these welds, VYNPC conservatively -=ad that all relevant indications were through-wall cracks.

VYNPC's flew evaluation of the H5 weld indicated that the H5 weld has sufficient structural margin to justify one more cycle of plant operation. In evaluating the H5

weld, VYNPC conservatively assumed that all uncracked 6

areas and all areas with indications less than 0.013 m (0.5 in) deep ca=*=iaad cracks 0.013 m (0.5 in) in depth.  ;

.VYNPC also conservatively assumed that areas not inspected contained continuous through-wall cracks, and that any relevant indications with crack depths greater than 0.013 m (0.5 in) were also through-wall in nature.

, 'Ibe flaw evaluations also included conservative allowances for crack growth and NDE uncertainties.

5 VYNPC submitted the results of its inspections and evaluations for review by the staff on April 20, 1995 (Ref.12). VYNPC's submittals included a consequence analysis (safety analysis) of the VY core shroud.

The staff reviewed VYNPC's examination results, flaw evaluations and consequence analyses regarding the VY I core shroud. The staff determined that the VY core shreud has sufficient remaining structural margins to justify one additional cycle of operation. The staff issued its SER reganting the structural integrity of the VY core shroud on April 27,1995 (Ref.13).

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5-3 NUREG-1544

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6 INDUSTRY EFFORTS TO ADDRESS THE IGSCC ISSUE i I

6.1 G-ic Anoroach Takaa to The plants that have experienced the most extensive Resolve the IGSCC Issue cracking have been operated for longer than 8 years and had histories of moderate to high coolant conductivities 1GSCC in BWR internal components is a long-term when averaged over the first 5 cycles of operation.  !

problem. As BWRs begin to age, the number ofIGSCC i incidents in BWR internal components is expected to  % BWROG evaluation indicated that the structural 1

increase. For this reason, the NRC has encouraged the margins for the plants with the most susceptible core BWR industry to take a conservative, long-term shrouds would be maintained for at least one additional approach to resolve the issue ofIGSCC in BWR internal cycle of operation at current conductivity levels. b components. De approach involves the following steps: BWROG concluded that it was unlikely that any i development of cracking would fail to satisfy the safety

- (1) direct interactmn between the NRC and the margins specified in Section XI of the ASME Code. I industry organiaanons, namely the Boiling Water Howevsr, because of uncertainties in the assumptions Reactor Owners Group (BWROG) and the Boiling used in the safety evaluation, such an occurrence could l Water Reactor Vennels and laternals Project not be ruled out. The NRC issued GL 94-03, in part.

- (BWRVIP) "to ase .rtain the likelyhood of such an occunence and to t b appropriate corrective action (s)" as necessary. Both (2) NRC annenamnet of generic guidelines established the BWROG and individuallicensees have indicated that by the BWR anJuntry organizations repairs would be implemented for cases in which it is uncertain that ASME Code margins could be met.

(3) NRC asseaamesW of plant specific actions on an individual hasse Revision 1 of the BWROG's submittal dated i July 13, 1994, was received on August 5,1994 The important aspect of this approach is that it is (Ref.16), along with a response (Ref.17) to a request proacriw rather than reersiw, since it encourages the for additional information (RAI) that the NRC forwarded i industry to deveenp and implement appropriate to the BWROG on May 12,1994 (Ref.18). The '

uupections programs akeg with predetermined BWROG's submittal of August 5,1994, categorized  ;

acceptance cneens and repair methods. Effective BWR core shrouds into seven IGSCC susceptibility  !

inspection programs wdl enable licensees to detect groups for ranking purposes. . Dese susceptibility cracking before se becomes a safety concern, and rankings were established to aid the individual BWR predetermined ==paaare enteria and repair methods utilities in their efforts to address the criteria establishej i will ensure optimal use of industry and NRC resources. in GL 94-03.

6.2 Efforts by the Boiline Water Reactor Owners Groun 6.3 Establishment of the Boiline Water Reactor i Vessel and Internals Proiect De BWROG submitted its criteria for evaluating BWR core shrouds in a letter to the NRC dated April 5,1994 In a meeting on June 28,1994, the BWROG informed (Ref. 14). De inspection strategy detailed in the the staff that a new industry organization, the Boiling BWROG suport focuses on a ranking system that bases Water Reactor Vessel and Internals Project (BWRVIP) a plant's IGSCC susceptibility according to its age, had been established solely to address the issue of age- ,

construction materials, and reactor coolant conductivity related degradation of BWR internal components. The 1 level. De BWROG then arvt=8 arf and refined its BWRVIP comprises five subcommittees: (1)lategration, susceptibility ranlungs, which were forwarded to the (2) Inspection, (3) Assessment, (4) Mitigation, and (5)

NRC in a submittal dated July 13,1994 (Ref.15). In Repair. Each subcommittee is chaired by both a top this submittal, the BWROG, in conjunction with CE, . executive from one of the BWROG member utilities and provided an evalusten of core shroud cracking o'merved a engineering staff member from the industry. This in domestic BWRs that had previously been inapected. organisation is designed to ensure that the BWRVIP's 6-1 NUREG-1544

efforts are reviewed on both the achaical n and executive which would ensure each weld's integrity during power levels, and to encourage w" A industry w , -= operation (Ref.19). The staff informed the BWRVIP of BWRVIP guadelines, criteria, and methods. To date, that this " minimum ligament inspection scope" was too individual BWROG ===hars have shown widespread narrow to give an accurate indication of cracking in a support for the BWRVIP's e% ts and activities, core shroud (Refs. 20 and 21). In Revision 1 to the

'BWR Core Shroud Inspection and Evaluation On hr ==he s 2,1994,' the BWRVIP submitted the Guidelines," dated April 21,1995 (Ref. 22), the

'BWR Core Shroud laspection and Evaluation BWRVIP amended its earlier recommendations by Guadehmes* to the NRC (Ref.19). %ese guadelines recomunending that inspection scope for Category 'C' m' ^ ' and ' ' the information regarding core shroud inspectica scopes and flaw evaluations core shrouds cover 100% of the accessible areas of circumferential welds H1-H7 (through H8 for BWR-2 contained in the BWROG genene safety ===== ant of designs). The staff concluded that the BWRVIP's August 5,1994, la sununary, the 'BWR Core Shroud alternate inspection scope recommendation for lampace== and Evaluation Guidelines' reduced the Category "C" shrouds was acceptable (Ref. 23). The number of susceptibility -: " .M from seven to three. NRC issued its SERs regarding the 'BWR Core Shroud De factors caaninianed in fornung the categones Inspection and Evaluation Guidelines, " Revision 0 and included hot operating time (until next refueling outage), Revision 1, on December 28,1994, and June 16,1995, mean reactor coolant conductivity values when averaged respectively. (Refs. 21 and 23).

over the first five operatsag cycles, carbon cantants of the cosa sluoud constructice matenals (Type 304 he BWRVIP has committed to submit additional se==l== steel vs. Type 304L stainless steel), and conespondence regarding BWR core shroud and internal ma*had= of fabncation. All operating BWR plants were components in the future. The proposed submittals will then grouped into one of tlues categones ("A," "B,* or provide an integrated safety assessment of the issue, re-

"C") based on the potential of their internal components mspection scopes and acceptance criteria, and mitigation to develop IGSCC and on previous field inspection measures, as well as changes to the core shroud repair exponence. '". ,i "A." "B," and "C" are criteria. A submittal from the BWRVIP Repair described in more depth in Table 6.2-1. Plant-specific Technical Subcommittee regarding repair options was data regarding the BWRVIP rankings are provided in received by the staff on August 18,1994 (Ref. 24).

Appendix B of this report. However, it should be noted his submittal provided information regarding suggested that plant categonantions may change as plants accrue criteria for the evaluation of licensee repair options.

operating time. The staff issued its evaluation of the BWRVIP repair design criteria on September 29, 1994 (Ref. 25).

De BWRVIP 'BWR Core Shroud Inspection and Figure 6.3-1 provides an example of a typical core Evaluation Guidaliaan" also recoau=aadad inspection shroud modification (repair) design that has been

=ahadata= and scopes based on the susceptibility rankings submitted to the staff for review. The staff will continue of the plants. De BWRVIP ' ;* guidelines to review core shroud modification design submittals on provided licensees with Category "A" plants the option a case-by-case basis.

of postponing core shroud inspections until eight cumulative years at power had elapsed at their facilities.

De BWRVIP reco==aadad that licensees with 6.4 Activities of the General Electric Comnany Category "B" plants perform limited VT-1 or UT iamparaman of their core shrouds at the next plant GE initially reported the cracking found at KKM in refushag outage. For licaa- with Category "C" RICSIL 054. On October 4,1993, GE issued Services pleets, the BWRVIP recomunaadad VT-1 or UT Information Letter (SIL) 0572, Rev.1 (Ref. 26), to

==paceiaan of welds H1 through H7 (through H8 for incorporate domestic data on shroud cracking, and to BWR-2 plants) at the next refueling outage. update recommendations for mspecting BWR core shrouds.

De BWRVIP did not initially recommend 100%

'-;%= of all accessible circumferential weld areas. In SIL 0572, Rev. 1. GE recommended that BWR lassmed, the BWRVIP initially stated that weld coverages licensees perform visual or ultrasonic inspections of their only had to be - _ , ? -- "ve to the extent they proved core shrouds for a statistically significant sample sire of the existance of sufficiently long, unflawed ligaments accessible welds and associated HAZs. GE also NUREG-1544 6-2 I

recommended that the inspections be performed after six notches and realistic IGSCC defects . Qualification of effective full-power years (EFPY) if the shroud is UT techniques is normally conducted at the EPRI NDE fabncated from normal carbon content austenitic Center in Charlotte, North Carolina. EPRI is also stainless steel (0.03% to 0.08% C), or after 8 EFPY if currently investigating whether or not eddy current the shroud is fabiicated from austenitic stainless steel of testing (ET) is an appropriate NDE method for BWR a low carbon content (< 0.03% C). GE also internal components. However, the NRC has not yet reconn=adad that licensees reinspect the shrouds at accepted ET for use on BWR internal components.

every refueling outage if cracking was observed, or every two outages if cracking was not observed. No guidance was given concerning structural integrity or  !

repair.

Metallurgical aspects of cracks in core shrouds fabricated from Type 304L stainless steels were also discussed in GE RICSIL 068, Rev. 1, dated April 14,1994 (Ref. 27). On May 6,1994, GE issued RICSIL 068, Rev. 2 (Ref. 28) to supplement RICSIL 068, Rev.1, and to update the lessons learned from core shroud visual and ultrasonic examination data oflow-carbon *L"-grade stainless steel core shrouds. In RICSIL 068, Rev. 2, GE also provided cautions about the adequacy of visual examination procedures, and ,

discussed whether ultrasonic examination methods are l preferable to visual examinatior. methods under certain ,

circumstances. I 6.5 Activities of the Electric Power R-ch Institute Bacausa core shroud examinations involve c. complex, j detailed set of activities, they must be plamw! well in

advance in order to be effective. A licence must first j de.

; bc which core shroud welds must be included in the mapection scope to provide for a sufficient

===a==amat of the core shroud, and thea must determine which NDE methods are best suited for these examinations. Core shroud NDEs normally involve manipulation of complex instruments by utility NDE and engineenng staff members.

i

'The BWR industry has contracted with the Electric

! Power Research lastitute (EPRI) to assist industry

licensees in implementing NDE examinations of their j BWR core shrouds and other internal components.

EPRI's efforts have included the design of a series of core shroud mockups, that can be used to qualify the UT

=cmaning equipment. These mockups are designed to contain electrodischarge machined (EDM) notches of known length and depth, and realistic IGSCC defects.

Qualification of the UT exanunation techmques car, then be accomplished by comparing the results of UT analyses to the known lengths and depths of the EDM 6-3 NUREG-1544 4

\

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1 Table 6.2-1 BWRVIP Susceptibility Rankings l and Core Shroud Inspection Recommendations' CATEGORY INSPECTION PLANT CHARACTERISTICS PLANTS RECOMMENDATIONS

'A' No inspection Plants with 304 SS shrouds, None necessary at < 6 years hot operating time, and l this time. avg. conductivities s 0.030 S/cm (0.030 pmhos/cm) during the first five cycles of operation. I Plants with 304L SS shrouds, Clinton, Fermi 2, Perry,

< 8 years hot operating time, and Hope Creek, Limerick 2, avg. conductivities s 0.030 S/cm Nine Mile Point 2, (0.030 pmhos/cm) during the first Washington Nuclear Plant 2, five cycles of operation. River Bend

'B' Limited inspection: Plants with 304L SS shrouds, Grand Gulf, top guide support ring, 2 8 years hot operating time, and Lasalle ! & 2, core support ring, and avg. conductivities s 0.030 pS/cm Limerick 1, mid shroud shell (0.030 mhos/cm) during the first Susquehanna 1 & 2 .

circumferential welds; five cycles of operation. )

also the bimetallic weld if accessible.

• C' Comprehensive Inspection: Plants with 304 SS shrouds and Shrouds - weld olate rines 1 circumferential shroud a 6 years hot operating time, Brunswick 1 & 2, welds Hi - H7 regardless of conductivity. Dresden 2 & 3, (and 118 for BWR-2s) FitzPatrick, Hatch 1, Millstone 1 Oyster Creek.

Nine Mile Point 1, Pilgrim, Quad Cities 1 & 2 Shrouds - forced rines Browns Ferry 1,2, & 3, Peach Bottom 2 & 3, Vermont Yankee, Monticello, Cooper Plants with 304L SS shrouds, Duane Arnold, Hatch 2 2 8 years hot operating time, and avg. conductivities > 0.030 S/cm (0.030 pmhos/cm) during the first five cycles of operation.

NOTES: 1. Medi6ed inna Table 3.1, 'BWR Core Shroud lagetties and Ev=6=*= Guidehmes." (Ref.19)

ABBREVIATIONS: 304 SS - Type 304 *===t=== Steel (Nonnal Canbes Content) 304L SS - Type 304L Stanniens Steel om Carbes Comtee0  !

pS/cus - Unit of conductivity is unicroSieuseen per centimeter l NUREG-1544 6-4 l

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Figure 6.3-1 Typical Modification Design Proposed for Repair of BWR Core Shrouds 6-5 NUREG-1544

(

)

, 7 GL 94-03, 1

"INTERGRANULAR STRESS CORROSION CRACKING OF CORE SHROUDS IN BOILING WATER REACTORS" e 7.1 Contant of GL 94-03 The NRC staff also requested that licensee's submit,

! under oath and affirmation, no later than 3 months On July 25, 1994, the NRC issuul GL 94-03, before inspectag or repairing of their core shrouds, the l "Intergranular Stress Corrosion Cracking of Core scope of the planned core shroud inspections and their l Shrouds in Boiling Water Reactors," to all BWR plans for evaluating and/or repairing their core shmuds l

! lic=- (with the exception of Big Rock Point, which based on mapection results. The NRC staff further I

! does not have a core shroud). The NRC staff requested requested that licensee's submit, under oath or j in GL 94 03 that licensee's take the following actions affirmation, their core shroud inspection results within j

, - with respect to their core shrouds: 30 days of completing their shroud examinations.

2 (1) Inspect BWR core shrouds no later than the j plant's next refueling outage. 7.2 C:.mic A:=- ^ of the M=Ws i

Resnonses to GL 94-03 (2) Perform matenals related and plant-specific safety analyses with respect to the core shrouds. h NRC's reviews covered the following items in the plant-specific responses to GL 94-03

} (3) Develop tore shroud inspection plans, which j address inspection of all core shroud welds and (1) schedules for inspection or repair of BWR core take into account the latest available technology shrouds j developed by the industry for inspection of BWR

internal components. (2) safety assessments based on postulated core i shroud failures

{ (4) Develop plans for core shroud evaluation and/or i repair. (3) scopes of BWR core shroud inspections j 4

The staff also reco===rtad in GL 94-03 that licensees (4) plant-specific inspection results work closely with the BWROG to address the issue of j 1GSCC of BWR internal components. (5) core shroud flaw evaluations, as appropriate

, ]

! i The NRC staff requested that licensees submit, under (6) core shroud repairs, as appropriate

! oath or affirmation, the following information ini j response to GL 94-03 within 30 days from the date of To facilitate its reviews, the NRC grouped the industry

issuance

core shrouds according to their relative susceptibility to IGSCC, as ranked by the BWRVIP Technical

. (1) a core shroud inspection schedule Subcommittee on Inspection. (The BWRVIP rankmgs

have been discussed in more depth in Section 6.3 of this j (2) a safety analysis supporting continued operation report.) The NRC then issued safety evaluation reports of the facility until inspections are conducted (SERs) for all BWRs in the industry, with the exception

of Big Rock Point, which does not have a core shroud.

(3) one or more drawings of the core shroud i configurations In order to simplify its task of determining whether or not individual BWR licensees could justify operation of

(4) a history of core shroud inspections completed to their units to the respective RFOs, the NRC performed

) date a generic assessment of the results of core shroud

inspections conducted before July 1995. h staff i

7-1 NUREG-1544 1

--m , e -

m ,+

l l

l-L deternu-t that no cases of 360* through-wall cracks occurred in any core shroud inspected before July 1995, and no BWR had exhibited any symptoms (power-to-flow mismatch) that would be indicative of bypass ,

leakage from a 360* through-wall crack. Furthermore, i the staff determined that, in all cases, sufficient ligaments remained in the core shreuds to provide a

• reasonable assurance that the structural integrity of the ,

shrouds would be maintained during the current plant i operating cycles. De staff also determined that, in all cases, the frequency of an initiating event which could challenge the_ structural integrity of a core shroud was i Iow. In addition. the uaff noted that only a short time remained before the scheduled RFOs when the licensees 1 would inspect or repair their respective core shrouds.

The NRC therefore concluded that, in all cases, the

-]'

licensees provided sutlicient technical bases to justify operation of thejr units to their next respective RFOs.  ;

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NUREG-1544 7-2 i

_ ~

8 PLANT-SPECIFIC ASSESSMENTS AND RESULTS OF CORE I

! SHROUD INSPECTIONS OR REPAIRS

~i 8.0 Overview core shroud was highly susceptible to IGSCC. In considering the plant-specific susceptibility factors for

his chapter provides the staff's assessments regarding PNPS, as well as the industry-wide inspection the ' plant-specific responses to GL 94-03 and a experience and the uncertainties in the residual stress discussion of the mdustry's inspection and repair profile for the PNPS shroud, the staff concurred with activities to date. Appendix B augments this discussion BECo's susceptibility assessment of the PNPS core i

by summarizing the plant-specific core shroud data, and shroud. The staff therefore concluded that significant providing an overview of pertinent information requested cracking in the PNPS core shroud could not be ruled j from licensees, concerning core shroud materials, out.

operation, and fabrication. De NRC found this information essential to its assessments of the BECO performed a preliminary plant-specific flaw

susceptibility of industry core shrouds to IGSCC. He evaluation of the PNPS core shroud as part of its JCO. I plant-specific core shroud summaries also include The results of the tiaw evaluation showed that only a 5 %

discussions of the plant-specific inspections and repairs remaining ligament of the PNPS shroud wall was needed 4,

performed by the industry through the end of to maintain the structural integrity of the core shroud September 1995, and a list of the corresponding SERs under all design con <'.nions. This evaluation was based and acknowledgement letters issued by the NRC in on a limit load ans'ysis (LLA) of the PNPS core shroud.

response to the industry's submittals to GL 94-03. BECo also used ne GE PLEDGE model to calculate a hypothetical crack growth rate for any postulated crack 4 in the PNPS core shroud during the remaining time in 4 8.1 Boston Edison Comnany the PNPS 1994-1995 operating cycle. However, since 4 the initial flaw size was not known, BEco used the

! 8.1.1 Assessment of the Response to GL 94-03 results of the BWROG's generic crack growth analysis for the Pilgrim Nuclear Power Station as a bounding analysis for the PNPS core shroud.

i l Boston Edison Company (BECo), the licensee for the The BWROG's generic. crack growth analysis was

Pilgrim Nuclear Power Station (PNPS), responded to W"ked using the worst crack depth measurements l GL 94-03 on August 27,1994 (Ref. 29). De licensee's associated with shroud inspections performed at i response included a schedule for inspection of the PNPS Brunswick Steam Electric Plant, Unit 1 (BR-1). De core shroud and a safety assessment to support continued results of the analysis predicted that the structural l operation of PNPS until the April 1995 RFO. In a integrity of the shroud wall would be maintained even '

, public meeting on October 4,1994, and in submittals with a postulated 360' crack extending up to a depth of dated October 13,1994 (Ref. 30), and October 28,1994 80% of the shroud wall thickness. In addition, since (Rif. 31), BECo provided additional information to hydrogen water chemistry was implemented at PNPS, support itsjustification for continued operation (JCO) of BECo assumed that postulated cracks in the shroud j PNPS until the April 1995 RFO. would grow less than 2.54 x 104 m (0.01 inch) during '

the remainder of the 1994-1995 operating cycle, even BECo reviewed the plant-specific susceptibility factors if a factor of 10 was applied to account for the

, regarding the PNPS core shroud. In submitting its uncertainties in the growth rate. BECo therefore  !

findings to the staff in its plant-specific response to concluded that a sufficient structural margin would be GL 94-03, BECo informed the NRC that the PNPS core maintained in the PNPS shroud to justify operation of  ;

shroud was fabricated from type 304 plate materials with PNPS until the April 1995 RFO.

4 carbon content typically in the range of 0.040-0.065 %.

1 BECo also stated that PNPS had operated in excess of BECO also performed a plant-specific safety assessment i 15 years, and that average reactor coolant conductivity . of the PNPS core shroud. BECo's intent was to over the first five years of operation was in excess of demonstrate that fuel geometry and core cooling would

0.300 pS/cm. BECo therefore concluded that the PNPS be maintained given the occurrence of a through-wall 8-1 NUREG-1544 4

e

, , _ _m-_,

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

E 4

circumferential weld failure concurrent with postulated core shroud in response to a postulated MSLB, RLB main steam line break (MSLB) or recirculation line (including acoustic and blowdown loads), MSLB plus break (RLB) conditions. BEco also performed the seismic event, and RLB plus seismic event, given a safety ==== ament - to determine whether any postulated through. wall failure of one of the shroud's circumferential weld failures would be detectable by circumferential wenb. During postulated MSLB control room operators during normal operstmas. conditions, BECo's calc 9ations demonstrated that the top guide would not lift above the fuel, indicating that BEco used the GE TRAC-G Model as the basic model oc lateral movement of the fuel would occur. The NRC i for determinmg the differential pressures across the staff concluded that this was reasonable; however, shroud head and shroud support during MSLB or RLB hac==a of inherent uncertainties in BEco's analysis conditions. BECo concluded that any leakage through a methods, the staff concluded that there was a small weld separation of 0.05 or more meters (two or more likelihood that the top guide would lift above the fuel inches) would be detectable during normal operations. assemblies during a postulated MSLB concurrent with a BECo also stated that the ability to maintain reactivity failure of one of the upper circumferential welds. Even control, fuel geometry, core cooling, and a refloodable if this were to occur, however, the staff concluded that volume was ensured with substantial margin, even safe shutdown of the reactor weuld be achieved by though degraded perfonnance was assumed in the manual initiation of the standby liquid control system design-basis event evaluations. On the basis of this l (SLCS).

assessment, BECo concluded that core shroud separation and/or displacement occurring during normal operations For postulated through-wall failures of circumferential l or anticipated events would have no effect on the shroud welds, the other initiating event of concern would >

primary safety functions of reactivity control and core be the RLB. BEco's calculations indicated that, during cooling, which are required to mitigate design basis a postulated failure of a lower circumferential weld events. concurrent with an RLB, the resulting blowdown forces would induce a momentary tipping of the shroud, but no The NRC staff used the results of the safety margin permanent lateral movement. For such shroud response, analyses of the BR-1 core shroud as its basis for the staff agreed that little core / annulus bypass would evalcating the .BEco safety ==== ament. h staff occur during the RLB, and that adequate core flooding considered the BR-1 core shroud to be as susceptible to would be maintained during the event. Modeling the IGSCC as the shroud at PNPS. Although significant behavior of shroud with a through-wall crack in response shroud cracking was identifiest at BR-1, the NRC staff to a postulated RLB is quite complex. Such modeling 4-*H that all welds had sufficient remaining involves making assumptions regarding crack surface ligaments to ensure adequate structural integrity of the frictional forces and competing forces in the vertical and shroud during normal operating, transient, and lateral directions. h staff therefore concluded that

-

"H design-basis accident conditions. The staff lateral motion of the shroud following an RLB could not therefore concluded that any postulated IGSCC in the he precluded.

PNPS core shroud should be bounded by that detected at other BWRs of similar design. "Iherefore, consulering The staff concluded that a lateral displacement of the that only a small remauung ligament is necessary to shroud less than the magnitude of the shroud thickness ensure core shroud structural integrity, and considering would result only in small bypass leakages. However, mdustry experience regarding shroud cracking, the staff the staff also concluded that any large lateral movement '

concluded that the PNPS core shroud should have of the PNPS shroud had the potential to open a sufficient romaanmg ligament for the remainder of the sQnificant leakage path through the shroud wall. In this operating cycle leading to the PNPS RFO in April 1995. case, the staff reached the following determinations:

' lie NRC also performed a qualitative ===a= nont of (1) No 360* through-wall core shroud cracking had BEco's F--- y = ==anamane for the PNPS shmud. been observed to date in any U.S. BWR that had

'Ihe staff found BECo's submittal to be a relatively performed a shroud inspection.

complete maa-amant behavior expected from the PNPS NUREG-1544 8-2

(2) All analyses performed by the licensee for PNPS (BR 1), responded to GL 94-03 on August 24, 1994 showed that even if cracking did exist in its (Ref. 38). The licensee's response included CP&L's shroud, sufficient ligamants would remam in the justification for continued operation (JCO) of the BR-1 shroud to ensure its structural integrity during reactor until RFO B110RI, which was scheduled to normat operating conditions, operational transients, commence in March 1995.

and postulated design basis events.

CP&L response also included the lacensse's review of (3) PNPS had not exhibited any of the symptoms the insterials, fabrication, and operational histories of (power-to-flow mismatch) caused by leakage the BR-1 core shroud. CP&L originally inspected the through a 360* through-wall shroud crack. BR-1 core shroud during RFO B109R1 in the sunumer of 1993, and submitted its inspection results and flaw (4) MSLB and RLB are low frequency events. evaluations to the staff in November 1993. De licensee's submittal indicated that CP&L performed UT (5) Only a short time remained until a repair would be inspections on accessible areas of welds H1-H6b.

implemented at PNPS. CP&L also performed VT-1 inspections on the accessible areas of welds H7, H8, and H9. (Section 5.2 In addition, the staff noted that BEco replaced summarizes the results of emaminations performad an the recirculation line piping, and operates the PNPS RCS BR-1 core shroud during the Summer 1993 RFO. Of with hydrogen water chemistry. These practices particular note was the report of a 360* crack at the H3 substantially lower the frequency of an RLB and weld.)

somewhat mitigate the potential for IGSCC to occur in the core shroud. H erefore, based on these CP&L implemented a permanent repair of welds H2 and deter =manions, and the operational availability of the H3 in the BR-1 core shroud to ensure the structural SLCS, the staff concluded that there was no undue risk integrity of the H2 and H3 welds during ---I , -

^

to the public health and safety for the approximate four operating cycles. His repair design, which involved month penod remaining in the PNPS 1994 - 1995 installing a series of clamps encompassmg the H2 and operating cycle, and that PNPS could safely continue to H3 welds, was submitted to the staffin November 1993.

be operated until the April 1995 RFO. The NRC issued On January 14,1994, the staffissued a safety evaluation its SER regarding BECo's response to GL 94-03 on (Ref. 6), concluding that the structural integrity of the November 28,1994 (Ref. 32) BR-1 core shroud would be maintamed for the remainder of operating cycle leading to RFO B110R1  :

8.1.2 Repair of the Pilgrim Core Shroud (Spring 1995).

i BEco opted to install a pre-emptive repair (modification) 8.2.2 Reinspection of the Brunswick Unit 1 af the PNPS core shroud during the April 1995 RFO. Core Shroud SECo's core shroud modification design was submitted to the NRC for review on January 16,1995 (Ref. 33), During RFO B110RI, CP&L remspected the BR-1 core and was later supplemented with additional information shroud to determine how much additional cracking of the on February 24,1995, March 14,1995, and April 17, BR-1 core shroud occurred during the plant's 10th j 1995 (Refs. 34-36). De NRC reviewed BECo's core operating cycle. On April 28,1995 CP&L submitted i shroud modification design, accepted the design in the results of the RFO B110R1 shroud examinations i I

April of 1995, and issued its SER regarding the Pilgrim (Ref. 39). %e staff will issue its SER regarding the core shroud modification on May 12,1995 (Ref. 37). reinspections of the BR-1 core shroud after fim.lizing its review of the BWRVIP's generic guidance regarding core shroud reinspections.

8.2. Carh Power and Mehe Caamaav 8.2.3 Assessment of the Response to GL 94-03 8.2.1 Ammanamant of the Response to GL 94-03 for Brunswick Unit 2 for Brunswick Unit 1 CP&L, the licensee for the Brunswick Steam Electric The Carolina Power and Light Company (CP&L), the Plant, Unit 2 (BR-2), responded to GL 94-03 on licensee for the Brunswick Steam Electric Plant, Unit 1 August 24,1994 (Ref. 38). This response included l

8-3 NUREG-1544 )

l l

CP&L's justification for continued operation of BR-2 H6a welds, and included these flaw evaluations in its until RFO B212R1, as well as CP&L's review of response to GL 94-03. The evaluations of these shroud materials, fabrication, and operational histories of the i welds indicated that the shroud would meet structural  ;

BR-2 core shroud. The staff assessed the materials, margins for all circumferential welds for the remainder  ;

fabrication, and operational histories of the BR-2 core of operating cycle No.12. He NRC staff concluded I shroud and found them to be similar to those of BR 1. that CP&L's previous inspection results and evaluations ne NRC staff therefore concurred with CP&L's justified continued operation of BR-2 for the current assessment that the BR-2 core shroud is highly operating cycle (Ref. 40). Reinspection of the BR-2 susceptible to IOSCC.

core shroud is scheduled for RFO B212R1 in the- '

Summer of 1996.

CP&L originally inspected the BR-2 core shroud during RFO B211R1 in the Spring of 1994. CP&L did not include the H3 weld in its inspection scope during the 8.3 Commonwealth Edison Categorv 'C" Pla=4 j outage. Instead, CP&L implemented a ja... ast repair of shroud welds H2 and H3 in the same manner used at 8.3.1 Assessment of the Response to GL 94-03  :

BR-1 during RFO B109RI. CP&L also performed a 40- for Dresden Unit 3 and Quad Cities Unit I to S0-percent inspection of the H2 weld before the repair ,,

to confirm some indications at H2 that were recorded Comed performed comprehensive examinations of the ,

during an earlier maintenance outage. Other inspections core shrouds at Dresden Unit 3 (DR-3) and Quad Cities of the BR-2 shroud included (a) visual inspection (VT-1) Unit I (QC-1) during May of 1994 (Ref. 41). The of 18 percent of the H1 weld, (b) ultrasonic testing (UT) licensee's UT results indicated the presence of of 78 percent of the H4 weld from the outside diameter . significant cracking at the H5 locations of the DR-3 and (OD) of the shroud, (c) VT-1 of 93 percent of weld HS QC-1 core shrouds. Comed's inspection results from the inner diameter (ID) and 30 percent from the confirmed that the DR-3 and QC-1 core shrouds are OD, and (d) VT-1 of 11 percent of welds H6a, H6b, )

highly susceptible to IGSCC. Comed pufu. d flaw i and H7 from the OD (all percentages relative to the total evaluations and integrated safety assessments of the circumferential length of the welds). CP&L did not DR-3 and QC-1 core shrouds, and submitted them to the inspect welds H8 and H9 based on the acceptable results staff for review (Refs. 42 and 43). {

of the VT 1 examinations of the corresponding welds in ,

the BR-l core shroud. Comed performed its flaw evaluatmos of the DR-3 and l

QC-1 core shrouds in accordance with the flaw ,

he BR-2 core shroud inspection results identified 23 evaluation guidelines and acceptance criteria specified in circumferential indications at weld H4, seven Section XI of the ASME Code. Section 5.3 of this circumferential indications at weld H5, and one report discusses the Comed inspection results and flew circumferential indication at weld H6a. The longest evaluations in greater depth. He staff reviewed the indications were 0.35 m (13.6 in) in length at weld H4, Comed flaw evaluations and safety assessments, and 0.30 m (11 in) in length at weld H5, and 0.038 m concluded that sufficient ligaments remained in the DR-3 '

(1.5 in)in length at weld H6a Rese inspection results and QC-1 core shrouds to ensure their integrity for an confirm that the BR-2 core shroud is highly susceptible additional 15 months of service. The NRC issued its to IGSCC. SER regarding the operability of the DR-3 and QC,-l {

core shrouds on July 21,1994 (Ref. 9). j CP&L's previous inspections of the BR-2 core shroud l

. were not as comprehensive as the mapections of the 8.3.2 Assessment of the Responses to GL 94-03 for j BR-1 shroud in 1993. However, the BR-2 core shroud Dresden Unit 2 and Quad Cities Unit 2 anspections were performed before the issuance of GL 94-03. De inspection scope was sufficient to assess . In its response to GL 94-03 (Ref. 41), ComaEd submitted the current status of the BR-2 core shroud with regard to its review of the materials, fabrication and operational the presence ofIGSCC. Although the BR-2 core shroud histories of the core shrouds at Dresden Unit 2 (DR-2) is considered highly susceptible to IGSCC, CP&L had and the Quad Cities Unit 2 (QC-2) After reviewing this gathered considerable inspection data regarding the information, the NRC concluded that the DR-2 and current condition of the core shroud. The licensee QC-2 core shrouds are susceptible to IGSCC, and that completed evaluations of the flaws in the H4, H5, and significant cracking of the DR-2 and QC-2 core shrouds NUREG-1544 8-4 6

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'l i l

could not be ruled out. With reg.rd to the QC-2 core shroud, the NRC deternuned that QC-2 had been on-line for Comed performed plant-specific safety assessments of approximately the same amount of time as has QC-1, ,

the DR 2 and QC-2 core shrouds in order to justify and had a reactor coolant water chemistry history similar contmued operation of the DR-2 and QC-2 plants until to that of QC-1. The NRC therefore concluded that the their respective 1995 RFOs. Comed assumed that as-found conditions and determinations given in the cracks could potentially initiate in the DR-2 and QC-2 staff's SER regarding the cracking found at QC-1 and core shrouds after three effective full power years DR-3 (Ref. 9) would bound any cracking that could (EPPY) of operation. Comed's structural integrity potentially occur in the QC-2 and DR-2 core shrouds.  ;

calculation for the DR-2 core shroud resulted in a In addition, the NRC concluded that the DR-2 and QC-2 '

bounding crack depth of 0.016 m (0.64 in) for the DR-2 core shrouds should have sufficient remaining ligaments core shroud. Comed's calculation also indicated that to justify continued operation of the DR-2 and QC-2

~9% remaining shroud ligament would be necessary to units to their respective summer and spring 1995 RFOs, maintain the structural sategrity of the DR-2 core and that operation of these units to the 1995 RFOs i

shroud, even under fauhad conditions (the most severe should not adversely affect to the health and safety of operating conditions for tlw plant). In performing these the public (Ref 44).

calculations, Comed used what they considered realistic crack growth rates, an desernuned from results of the GE 8.3.3 Repairs of the Dresden, Units 2 and 3, and 7 PLEDGE model. Comed concluded that the remainmg Quad Cities, Units 2 and 3, Core Shrouds  ;

ligaments projected for the DR 2 shroud welds would provide considerably grenser margin than that required On January 16, 1995 and May 24,1995, Comed i by Section XI of the A5ME Code. submitted its modification designs for the QC-1 & QC-2, and DR-2 & DR-3 core shrouds, respectively (Refs 45 Comed also stated that the wune crack depths measured and 46). Comed's design modification for the core  ;

ct QC-1 during the rnent inapections performed during shrouds involves installation of a series of GE designed .

the Spring RFO wtadd he huandmg for any postulated tie rod assemblies around the OD of the shreuds. The cracking in the QC 2 cure ehrtnad. Comed based this tie rod assemblies are designed to provide an alternative conclusion on the chicnasman that the water chemistry load bearing capability during the most severe normal conditions over the first 5 years of operation were operating, transient and postulated design basis accident '

similar for the QC 1 and QC-2 umts, and that QC-2 has conditions for the plants, given the occurrence of a 360*

operated at power for 2 ycer lens time than QC-1. through-wall failure of a circumferential shroud weld.  ;

The NRC reviewed and approved the modification

'Ibe DR 2 and QC 2 core shnmads are of a construction design for the QC-1 and QC-2 core shrouds on ,

similar to that of DR.3 and QC l core shrouds, which June 8,1995 (Ref. 47). The NRC evaluated and were both inspected in the spring of 1994, approved the core shroud modification design for DR-2 .

Consequently, the NRC annensed the DR-2 and QC-2 and DR-3 in the fall of 1995 (Oct.10,1995).8 core shrouds by benchmarking the plant-specific data i against data which was previously obtained through evaluations of the DR-3 and QC-1 core shrouds. The 8.4 General Public Utilities mapection results of the DR-3 and QC-1 core shrouds confirmed the existence of significant, circumferential 8.4.1 Assessment of the Response to GL 94-03 for the cracks associated with the H5 welds of the DR-3 and Oyster Creek Nuclear Generation Station QC-1 core shrouds. The NRC staff noted that the DR-2 unit had operated 2 years longer than the DR-3 reactor. General Public Utilities (GPU) submitted their response I However, based on plant-specific susceptibility criteria, to GL 94-03 on August 24,1994 (Ref. 48). 'lhe the NRC staff concluded that the DR-2 core shroud BWRVIP categorized the OCNGS shroud as being I appears less susceptible to IGSCC than the DR-3 shroud.  ;

The staff based this conclusion on the fact that the DR-2 3 Although the date of the staff's SER regarding the J reactor has been operating with hydrogen water Dresden Core Shroud Modification Design is outside '

chemistry for the past seveial operating cycles, and that the time frame scope of this report, the date is listed l Comed had reported more incidents of cracking in here to indicate that the staff did approve the design '

safety related components at DR 3 than at DR-2. for installation at the site.

8-5 NUREG-1544 uw -

u- _ w ---a _a-_ w_ m- --m-.-.- ?y --- r

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

l 4

highly susceptible to IGSCC, and rated the OCNGS design basis accident conditions, given the occurrence of shroud as a Category 'C' shroud, based on a through-wall failure of a circumferential weld in the shroudmaterials and fabrication data, the number of plant HAT-1 core shroud. GPC installed the proposed years on-line, and the plant specific reactor coolant modification to the HAT-1 core shroud during the 1994 chemistry history. Upon reviewing the materials, RFO, which commenced on September 15,1994. (The i fabrication, and operational data regarding the OCNGS core shroud modification was implemented in lieu of core shroud, the staff concluded that the BWRVIP's comprehensive core shroud examinations.) ne amaa==aat of the OCNGS core shroud was appropriate modification involved the installation of a number of (Ref. 49). GE-designed tie rod assemblies placed symmetrically into the annulus between the reactor vessel wall and the 8.4.2 laspections and Repair of the core shroud wall.

Oyster Creek Core Shroud ,

8.5.3 Assessment of the Response to GL 94-03 GPU inspected the OCNGS core shroud during for Edwin 1. Hatch Unit 2 ,

RFO 15R which commenced on September 10, 1994.

b ex====sions of the OCNGS core shroud revealed Georgia Power Company (GPC), the licensee for the significant crackmg in the H4 weld. After completing Edwin 1. Hatch Nuclear Plant Unit 2 (HAT-2),

the UT examiamhons of the H4 weld, GPU decided to responded to GL 94-03 on August 24,1994 (Ref. 50),

modify the OCNGS core shroud (Ref.10). (Section 5.4 he licensee has indicated that they have scheduled a  ;

discusses the OCNGS core shroud examination results modification of the HAT-2 core shroud during the fall ,

and repair design in greater detail.) h staff reviewed 1995 refueling outage (RFO) in the same manner that GPU's core shroud modification design and accepted the was used to modify the HAT-1 shroud during the fall  ;

modification on Nov. 25,1994 (Ref.1I). 1994 RFO.

The BWRVIP categorized the HAT-2 shroud as a 8.5 Georma Power Comnany Category 'C' shroud. b NRC staff concluded that the BWRVIP assessment of the HAT-2 shroud was 8.5.1 Annan==aat of the Response to GL 94-03 appropriate (Ref. 51).

for Edwm I. Hatch Unit 1 .

GPC originally examined the HAT-2 core shroud during ,

Georgia Power Company (GPC), the licensee for Edwin the spring 1994 RFO. D ese inspections were

1. Hatch Nuclear Plant Unit 1 (HAT-1), responded to performed prior to issuance of GL 94-03. GPC GL 94 03 on August 24,1994 (Ref. 50). This response performed the UT examinations of the HAT-2 core included GPC's review of the materials, fabrication, and shroud using the GE O.D. Tracker UT Scanner, the GE  ;

operational histones (water chemistry and on-line years) SMART 2000 Data Acquisition / Analysis System, and a of the HAT-1 core abroud. GE designed motion control system. GPC also performed enhanced VT-1 examinations in accordance i h BWRVIP determined that the HAT-1 shroud is with the recommendations found in GE SIL 0572, highly susceptible to IGSCC, and rated it as a Revision 1. using a high-resolution camera capable of Category 'C' shroud. He staff determined that the resolving a 2.54 x 104 m (0.001 in) wire on a gray ,

BWRVIP amman==aat of the HAT-1 shroud was background. De scope of GPC's examinations included ]

appropnate, and concluded that significant cracking of 100-percent UT examinations (from the OD surface) of ,

the HAT-1 core shroud could not be ruled out (Ref. 51). the accessible areas of welds Hi - H4, and partial i Ami VT-1 inspections from the OD surface of 8.5.2 Repair of the Edwm 1. Hatch Unit 1 Core Shroud shroud welds H5, H6a, H6b, H7, and HB, commencing at the O' and 180" azimuthal locations.

On hpta-t- 2, 1994, GPC submitted a design modification for the HAT-1 core shroud (Ref. 52). In GPC identified IGSCC indications at several of the its SER of September 30,1994 (Ref. 53), the staff circumferential welds, including: (1) five indications at concluded that the HAT-1 core shroud design shroud weld Hl. the longest being 0.23 m (9 in) in modification would ensure the structural integrity of the length; (2) nine indications at shroud weld H2, the core shroud during normal, transient, and postulated longest being 4.04 m (159 in) in length (= 1/4 around NUREG-1544 8-6

l the circumference of the shroud at this location); 8.6 IES Utilities. Inc.

(3) eight indications at shroud weld H1, the longest being 0.43 m (17 in) in length; and (4) fifteen 8.6.1 Assessment of the Response to GL 94-03 indications at ehroud weld H4, the longest indication for the Duane Arnold Energy Center s being 0.30 m (1I in) in length. No relevant IGSCC  !

indications were ' discovered by GPC in the areas IES Utilities (IES), the licensee for the Duane Arnold inspected on the H5 - H8 welds. Energy Center (DAEC), v.id to GL 94-03 on l August 24,1994 (Ref. 56). 'Ihe IES response included ]

GPC performed a flaw evaluation of the HI - H4 weld a schedule for inspecting the DAEC core shroud, J indications in order to show that the HAT-2 core shroud justification supporting continued operation (JCO) of the '

would maintain its atmetural margins for the next plant, a description of the shroud, and a discussion of HAT-2 operating cycle (Ref. 50). GPC included this past core shroud mapection results. IES based its JCO flaw evaluation as part their response to GL 94-03. on the susceptibility of core shroud material to IGSCC, GPC's flaw evaluation was used to calculate the and on the absence of observed cracking in previous maximum allowable flaw lengths hamad on the most limited inspections of the shroud.

conservative stress magnitudes in the core shroud. Both LLA and LEFM methods were used for the analysis of b core shroud at DAEC is constructed from type 304L j the H4 weld indications. Only LLA was performed for stainless steel h higher resistance to IGSCC of this '

evaluation of the HI, H2, and H3 welds, where the material compared with type 304 stainless steel decreases neutron fluence levels are lower. GPC's flaw the likelihood that extensive cracking will be observed in svaluations included adjustments to account for crack the DAEC core shroud. The average conductivity value proximities, crack growth, and NDE examination of the DAEC reactor coolant during the first 5 years of uncertainties. operation was in the moderate range of conductivities for l the industry. Although the NRC anticipated that some j h staff reviewed GPC's flaw evaluation of the HAT-2 cracks may have initiated during the early years of core shroud and determined that the evaluation used operation, the staff concluded that the low carbon conservative methods to deternune safety margins content of the DAEC shroud materials would tend to remauung in the HAT-2 shroud. b staff also inhibit early IGSCC initiation.

determmed that the calculated safety margins were within the required values as specified in Section XI of During RFO No.11 in 1990, IES performed limited the ASME Code. h staff therefore concluded that the inspections of the DAEC core shroud at accessible HAT-2 core shroud had sufficient ligament to justify portions of several vertical shroud welds and at a single operation of the HAT-2 reactor for the remainder of the horizontal weld in the beltline region of the core. 'Ihese operating cycle (Ref. 51). inspections did not identify any indications of cracking in the DAEC core shroud. IES performed additional 8.5.4 Repair of the Edwin L Hatch Unit 2 Core Shroud inspections of horizontal and vertical shroud welds in 1993 (RFO No.12). These inspections again did not On July 3,1995, GPC submitted a proposed design identify any flaw indications in the DAEC core shroud modification for the HAT-2 core shroud (Ref. 54). 'the However, because of the limited swpe of past

. design modification involved installing a number of GE- inspections, the NRC concluded that the possibility of designed tie rod assemblies symmetrically about the OD significant cracks in the shroud could not be discounted.

of the core shroud wall. In its SER dated September 25,1995 (Ref. 55), the staff concluded that in order to assess the IES JCO further, the staff applied the HAT-2 core shroud modification would ensure the its generic core shroud assessment (i.e., the assessment structural integrity of the cose shroud during normal, of core shrouds that were mspected before GL 94-03 transient, and postulated design basis accident was issued, as discussed in Section 7.2). In this case, conditions, given the occurrence of a through-wall the staff determined that its conclusions in the generic f;ilure of a circumferential weld in the HAT-2 core safety assessment were applicable to the DAEC core shroud. GPC installed the HAT-2 tie rod assemblica shroud.

during the RFO in October 1995.

8-7 NUREG-1544

In its generic core shroud assessment, the staff NPPD performed a crack growth analysis using the GE concluded that the most highly susceptible core shrouds PLEDGE Model to justify operation of the CNS unit could contain cracks up to 80-percent of the shroud wall until the October 1995 RFO. NPPD's analysis indicated thickness, and still satisfy the applicable safety margin that most of the growth of an IGSCC-initiated crack requirements of Section XI of the ASME Code. The would occur during the early portions of plant life, but NRC concluded that the conditions needed for initiation would be significantly reduced after the first five cycles.

and growth of IGSCC in the DAEC core shroud would NPPD's analysis also indicated that only a 7-percent be bounded by the conditions at the most highly shroud ligament was necessary to ensure the structural susceptible BWRs. Consequently, any postulated integrity of the shroud.

IGSCC in the DAEL cere shroud should be bounded by the most severe cracking identified at those BWRs. The The NRC did not accept the results of NPPD's crack staff therefore concluded that, while significant cracking growth analysis based on GE PLEDGE Model, lastead, could not be entirely ruled out, the DAEC core shroud based on the information provided by NPPD, the staff should retain adequate stmetural margin to justify safe concluded that the material and water chemistry operation of DAEC until the February 1995 RFO evaluations indicate that IGSCC initiation in the (Ref.56). horizontal welds could occur in the CNS shroud welds.

Since the initial depth of a postulated crack at CNS 8.6.2 Inspection of the Duane Arnold Core Shroud could not dermitely be established, staff determined that its generic core shroud assessment was appropriate for IES performed comprehensive inspections of the DAEC its evaluation of the CNS core shroud. M staff core shroud during the winter 1995 RFO. % scope of concluded in its generic core shroud assessment that the IES's examinations of the DAEC core shroud covered most highly susceptible core shrouds could contain welds HI - H7. IES performed the examinations using cracks extending up to 80-percent into the shroud wall the GE OD Tracker System. His system includes 45' thickness, and still satisfy the minimum required safety shear wave and 60' longitudinalwave transducers. IES margins (as specified in Section XI of the ASME Core) also used UT creeping wave methods for additional during the operating cycle.

surface examinations of core shroud welds Hi - H6A.

All examinations of the DAEC core shroud .were The staff concluded that a generically determined crack negative for relevant indications. The results of the depth of this size was conseivative in relation to the DAEC shroud inspectionsjustify operation of the DAEC thickness of the CNS shroud hacanoe the water core shroud for the current operating cycle. chemistry at CNS was significantly better than that assumed in the generic evaluation. Even if this limiting crack depth value was assumed, the staff deternuned that 8.7 Nebraska Public Power Distligt sufficient stmetural margin would be maintained for the remainder of the current CNS operating cycle since the 8.7.1 Assessment of the Response to GL 94-03 predicted crack growth for the current cycle was for the Cooper Nuclear Station expected to be small. Based on these =====nants, the staff determined that the proposed schedule for Nebraska Public Power District (NPPD), the licensee for inspection or pre-emptive repair of the CNS core shroud Cooper Nuclear Station (CNS), responded to GL 94-03 was acceptable, and that CNS could safely continue.to on August 26,1994 (Ref. 57). NPPD's response be operated until the October 1995 RFO (Ref. 59).

included its review of the materials, fabrication processes and operational histories (water chemistry and 8.7.2 Inspection Scope for the Cooper Core Shroud on-line years) of the CNS core shroud De BWRVIP categorized the CNS core shroud as a Category 'C' By letter dated July 14, 1995, NPPD submitted its shroud. N NRC staff reviewed the materials, inspection plan for the CNS core shroud (Ref 60),

fabrication, and operational history information provided NPPD's inspection scope includes UT exm====tions of by NPPD, . and determined that the BWRVIP's the accessible portions of circumferential welds H 1-H7.

susceptibility assessment of the CNS core shroud was This scope is consistent with the latest inspection criteria appropriate. N staff therefore concluded that cracking established by the BWRVIP for Category "C" core in the CNS core shroud could not be ruled out. shrouds (Ref.22). & staff accepted NPPD's NUREG-1544 8-8

_ __ __ _ _ _ _ _ . _ _ _ m - . _ _. _ _.___ . _ . _ _ _ _

i' 1

4 4

inspection. scope for the CNS core shroud on However, upoo finai review, NMPC decided to install

September 20,1995 (Ref. 61). the tie-rod assemblies in lieu of performing

{ comprehensive core shroud examinations. NMPC did not install the brackets. NMPC's omission of the

8.8 Nimmara Mohawk Power comorneian brackets in the design was based on the UT results of i Catenorv "C" Plants the H8 weld, which did not indicate the presence of any j flaws in the weld. The tie rod assemblies were designed i j 8.8.1 Ammanamant of the Response to GL 94-03 to provide a redundant load carrying capability for the i

! for Nine Mile Point Unit i NMP-1 shroud welds H1-H7. He staff reviewed the shroud modification hardware, considering structural,

, ne Niagara Mohawk Power Corporation (NMPC), the systems, materials, and fabrication factors. On the bases i licensee for Nine Mile Point Unit 1 (NMP-1), responded of that review, the staff concluded that the proposed 1 to GL 94-03 on August 23,1994 (Ref. 62). NMPC's modification of the NMP-1 core shroud was acceptable l response included a safety assessment to justify for implementation, and should ensure the integrity of j continued operation of NMP-1 until the core shroud the NMP-1 core shroud during subsequent operating 1 inspection scheduled for the February 1995 RFO. On cycles. The NRC issued its SER regarding the NMP-1 October 14, 1994, NMPC presented additional core shroud modification on March 31,1995 (Ref. 6,5),

informataan to the staff during a meeting at the NRC headquarters regarding their structural integrity j

===a=====e of weld H8. During the meeting, NMPC 8.9 Northeast Nuclear Enerev Comaany cleo showed the staff portions of their videotape of previous ~ inspections of weld H8. In its SER of 8.9.1 Assessment of the Response to GL 94-03 January 13,1995 (Ref. 63), the staff determined that its for the Millstone Unit 1 Core Shroud conclusions in the staff's generic assessment, as I previously discussed in Section 7.2 of this report, were Northeast Nuclear Energy Company (NNECO, a applicable to the NMP-1 core shroud. The staff subsidiary of Northeast Utilities), the licensee fo'r the therefore concluded that the presence of a remaining Millstone Unit 1, (MS-1), responded to GL 94-03 on ,

ligament in the NMP-1 core shroud, coupled with a low August 24,1994 (Ref. 66). However, this response j frequency of an initiating design-basis event and the omitted the detailed materials and fabrication history of l availability of the NMP-1 SLCS, provided a reasonable the MS-1 core shroud and an operational history of the I assurance that the NMP-1 core shroud would meet the MS-1 reactor. The licensee's basis for omitting this l applicable safety margins (specified in Section XI of the information was that such information was not needed to ASME Code) for the remamder of the operating cycle justify operation of the unit during the current operating

' leading to the Spring 1995 RFO. cycle (Cycle 15). Instead, NNECO provided the results of the core shroud inspections and flaw evaluations j 8.8.2 Repair of the Nine Mile Point performed during the Cycle 14 RFO (Winter 1994) as its l Unit 1 Core Shroud basis forjustifying continued operation of the MS-1 unit i during Operating Cycle 15. The staff agreed that this i NMPC submitted the NMP-1 core shroud modification was an acceptable basis for omitting the fabrication and design to the NRC on January 6,1995 (Ref. o4). operational history in the response to GL 94-03.  !

laitially, NMPC's scope for this modification poposed the installation of MPR-designed tie rod asscablies to Based on the staff's review of applicable plant specific i assume the loads acting on welds H1-H7, and a number data provided by the BWRVIP, the staff determmed that ef MPR-designed brackets to assume the vertical load the MS-1 core shroud is fabricated from a material  !

acting on weld H8. NMPC intended to examine the known to be susceptible to IGSCC. However, the staff H1-H8 shroud welds in accordance with the BWRVIP also determined that NNECO has typically kept the inspection criteria and to install the tie-rod assemblies impurities in MS-1 reactor coolant at levels well below and/or the brackets only if the inspections revealed that the industry norm. He staff therefore concluded that the rema==g ligaments in the shroud were not sufficient BWRVIP's categorization of the MS-1 core shroud as a l to meet the required ASME Code (Section XI) safety moderately susceptible Category 'C' shroud was margins, appropriate.

8-9 NUREG-1544

NNECO en==anad the MS-1 core shroud in accordance (October 1995). NNECO's inspection scope includes with the reco==aadations of GE SIL 572, Rev. I UT examinations of the accessible portions of the (Ref. 4), which contained the most up-to-date accessible portions of circumferential welds Hi-H7.

reco==aadations for performing shroud examinations at This scope in consistent with the latest inspection criteria the time of the ex==ia=* ions. b MS-1 core shroud established by the BWRVIP for Category "C" core VT-1 e======tions were performed from both the OD shrouds (Ref. 22). The staff accepted NNECO's and ID of all accessible shroud weld surfaces. NNECO inspection scope for the MS-1 core shroud on performed the VT-1 ex=====tions to a resolution of at August 11,1995 (Ref. 69).

least a 2.54 x 10 8m (0.001 in or 1 mil) wire. NNECO '

cleaned the weld surface areas by manual brushing before performing the en=nunations. In addition, 8.10 Northern States Power Company NNECO used a Westinghouse Model ETV 1250 black-and-white camera to inspect the welds and Westinghouse 8.10.1 Assessment of the Response to GL 94-03 for the twin quartz tungsten lamp assemblies to illuminate the Monticello Nuclear Generation Plant wolds. All of these choices were in accordance with GE's reco==aadations for performing enhanced VT-1 Northern States Power Company (NSP), the licensee for e===inations of core shroud welds. the Monticello Nuclear Generation Plant (MNGP),

responded to GL 94-03 on August 23,1994 (Ref. 70).

The staff noted that NNECO inspected the MS-1 core NSP's response included its review of the materials, shroud before GL 94-03 was issued. With the exception fabrication and operational histories (water chemistry.

of omitting the mapection of the H5A weld, the and on-line years) of the MNGP core staroud.

licensee's inspection scope for the Cycle 14 RFO shroud examinations agreed with the staff's position The BWRVIP has categorized the MNGP core shroud as

. --:-- -t exa-an=aion of 100% of the ace ==ible a moderately susceptible Category "C" shroud. The weld area of core shroud welds Hi-H7 (through weld NRC staff concluded that the BWRVIP's ranking of the H8 for GE BWR-2 designs). MNGP core shroud was appropriate hacanaa NSP used forged fabrication methods for manufacture of the core NNECO indicated that all fisw indications in the MS-1 shroud rings, and because NSP was able to maintain the core shroud were evaluated per Plant Nonconformance M NGP reactor coolant impurities at levels slightly lower Report # NCR 194-097. 'Ibe liccanee determined that than the industry averages (Ref 71).

all indications were acceptable for service during the current operating cycle. The staff did not formally 8.10.2 Inspection of the Monticello Core Shroud review the hc===aa's method of evaluating the flaw indications in the MS 1 core shroud. Nontheless, upon NSP completed comprehensive inspcetions of the MNGP reviewing the lic===aa*s inspection results, the staff core shroud during the September 1994 RFO. NSP detersoned that the size of the flaws found by the inspected 100 percent of the accessible areas of shroud licensee durmg the MS-1 shroud inspections were within welds HI-H5 using qualified UT examination methods, the screening criteria previously established by the and shroud welds H6, H8, and H9 using approved BWROG. 'Ibese evaluation methods are consistent with ct .ad VT-1 examination methods. NSP also the ==akad= used to evaluate the structural integrity of performed aah=aw VT-1 examinations of the H4 and the Brunswick Unit I core shroud, and are acceptable H5 welds to augment the UT examinations of those for use by the industry. The staff therefore concluded welds. 'lhese inspection were done in accordance with that sufficient structural margin exists in the MS-1 core the guidance provided by the BWRVIP.

shroud to justify operation of the MS-1 reactor during Operating Cycle 15 (Ref. 67). NSP did not examine the H7 shroud weld, as this weld area was inaccessible to both the UT tracker equipment 8.9.2 8 ';: d= Scope for the (as a result of obstruction from thejet pump assemblies)

Mill =eana Unit 1 Core Shroud and to the enhanced VT-1 camera (as a result of a weld backing bar). NSP also did not inspect the vertical By letter dated July 14,1995 -(Ref. 68), NNECO welds in the MNGP core shroud on the basis that submitted its scope for the MS-1 core shroud vertical welds were not necessary to maintain the asamiestaces scheduled for the Operating Cycle 15 RFO structural integrity of the core shroud. The staff agreed NUREG-1544 8-10 >

s 1

7 with the position that an mapection of vertical core coolant conductivities when averaged the initial 5 years shroud welds is m== y for core shrouds not of power operation.

modified by repairs. Herefore, the licensee's decision not to include these welds in the scope of the MNGP From a materials standpoint, the PB-2 and PB-3 core core shroud inspection was acceptable to the staff. shrouds were each constructed with ring segments fabricated from forged type 304 stainless steel, and shell On October 25,1994, NSP submitted the results of the segments fabricated from type 304 stainless steel plates.

MNGP shroud inspections performed during the fall Previous inspections of circumferential and vertical 1994 RFO (Ref. 72). NSP indicated that the flaw welds in the PB-3 core shroud revealed a crack of evaluations of the MNGP core shroud were poifvia_d moderate size (2.67 m or ~ 105 inches in length) in the in accordance with the criteria of the ASME Code lower HAZ of the shroud's H3 weld, and some lessSection XI and the flaw evaluation guidance developed significant crackmg at the H1 and H4 weld locations.

by the BWRVIP. NSP's examinations of the MNGP PECo therefore stated that any potential cracking of the core shroud revealed minor indications at welds PB-2 core shroud would be bounded by the amount of H2-H5. N maximum number of indications (five cracking in the PB-3 core shroud.

relevant flaw indications) occurred at weld H5. NSP's inspection results revealed that all flaw indications were b BWRVIP classified the PB-2 core shroud as a less than 0.25 m (10 in) in length, even after adjusting susceptible Category 'C' shroud. He staff concluded crack lengths to account for crack proximity that the BWRVIP's categorization of the PB-2 core relationships. All other UT and enhanced VT-1 shroud was appropriate, and that the PB-2 core shroud examinations performed by NSP were negative for should not be any more susceptible to IGSCC than the (

relevant indications. core shroud at PB-3 (Ref. 74).

De staff deteramed that NSP's inspection results were 8.11.2 Inspection of the Peach Bottom Unit 2 within the screening criteria previously established by Core Shroud the BWR Owners Group, and that NSP's evaluation '

methods were consistent with the methods used to By letter dated November 7,1994 (Ref. 75), PEco evaluate the structural integrity of the Brunswick Unit I resummarized the scope of the PB-2 core shroud (BR 1) core shroud (Ref. 6). W staff determined that inspections, and submitted the results of core shroud these flaw evaluation methods were acceptable for use examinations performed during the RFO 2R10. PEco by the industry. The staff also concluded that, based on performed the PB-2 shroud examinations using GE's the licensee's evaluation ofidentified cracking, sufficient Smart-2000 Data Acquisition System, OD Tracker and l structural margin remained in the MNGP core shroud to suction cup scaaners. The UT examinations used three justify operation of the MNGP reactor for the operating types of transducers: a 45* shear wave transducer, a 60*

cycle following the fall 1994 RFO (Ref. 71). longitudinal wave transducer, and a creepag wave transducer used to pick up surface indications. The  :

creeping wave transducer was not used on the H3 weld l 8.11 PMI lalaM= Electric Coma av because of equipment failure.

Cana-orv 'C" P!==*e i The scope of the UT examinations included all l 8.11.1 A=aamawat of the Response to GL 94-03 for the accessible portions of shroud welds H1-H7. His l Peach Bottom Atomic Power Station Unit 2 corresponded to approximately 33-percent coverage of  !

weld H1, 84% to 89% coverage of welds H2-H5, and l He Philadelphia Electric Company (PECo), the licensee 9% to 10% coverage of welds H6 and H7. PEco's

]

for the Peach Bottom Atomic Power Station Unit 2 inspections of welds H6 (the core support ring-to-lower 1 (PB-2), responded to GL 94-03 on August 24, 1994 shroud weld) and H7 (the lower shroud-to-shroud (Ref. 73). PECo used the results of previous inspections support cylinder weld) were conducted through of the Peach Bottom Atomic Power Station Unit 3 accessible areas of the access hole covers (AHCs).

(PB-3) as its basis for justifying operation of PB-2 until Interference fromjet pump assemblies, the reactor core, the RFO 2R10 (September 1994). PECo has operated and other internal components located at lower vessel the PB-2 and PB-3 reactors for approximately the same elevations limited access of the UT equipment to these number of years at power, and with similar reactor welds. PECo performed some additional anhanced VT-1 8-11 NUREG-1544 i

I j

l '

examinations of shroud weld H6 (-13 percent of the determined that it was acceptable. b staff therefore weld) to achieve some additional coverage.

concluded that PECo's evahiations of the PB 2 core shroud provided a reasonable assurance that the i

PECo identified the following relevant indications using structural integrity of the PB-2 shroud would be with the 45'S/60*L UT transducers:

i ASME Code safety margins and that PB-2 could be; 3

11 indications at the H1 weld, totalling 0.86 m safely operated for Operating Cycle No.11 (34 in), with the maximum length and maximum  ;

4 depth being 0.12 m (4.8 in) and 0.019 m 8.11.3 Assessment of the Response to GL 94-03 for the ;

(0.74 in), respectively (both at Indication #7) Peach Bottom Atomic Power Station Unit 3 l i

The Philadelphia Electric Company (PECo), the licensee ,

19 indications at the H3 weld, totalling 1.74 m (68.5 in), with the maximum length being for the Peach Bottom Atomic Power Station Unit (PBO), responded to GL 94-03 on August 24, 1994 '

O.22 m (8.75 in) at Indication #16 (Ref. 73). PECo completed an inspection of the PB-3 '

8 indications at the H4 weld, totalling 0.292 m core shroud during the previous Fall 1993 refueling ;

outage (RFO 3R9). PECo submitted the exm===83na '

(11.5 in), mth the maximum length being O.146 m f5.76 in) at Indication #4 results and assessment of the PB-3 core shroud to the NRC by letter dated March 14,1994 (Ref. 77).

1 indicatum at the H6 weld, 0.12 m (4.73 in) in

' length and o. Il m (0.45 in) in depth PECo reviewed h materials, fabrication and operational -

histories of the PB-3 core shroud and included this information in its response to GL 94-03. PEco PEco identified a munsw asunant of cracking at the H4 and H5 welda uusg UT creeping wave methods. determined that the mean initial 5 year conductivity of i the PB 3 reactor coolant was greater than the  !

Examinations of uwe ahnmal welds H2 and H7 were corresponding mean 5 year conductivity value for the negative for releast indecations, industry. In addition, PEco determined that the shell ;

PECo's inspectuin tramits were corapared to the initial portions of the PB-3 core shroud are fabricated from )

high carbon content Type 304 stainless steel patas. On screening cnteria ensahisahed as GENE 523-176-1293, these bases, both PECo and the BWRVIP has classified

" Evaluation and Ssecesung Critetis for the Peach Bottom Unit 2 Shroud * (Ret 76) If unxceptable, these results the PB-3 core shroud.as a susceptible Category "C" shroud.

were evaluated lov safety marpns using the LLA b staff concluded that the BWRVIP's susceptibility assessment was appropnate. This . !

methodology found in the 'BWR Core Shroud laspection -

conclusion was supported by the identification of i and Flaw Evaluation Gualelines* (Ref.19). This methodology appleen the most conservative loading moderate cracking during the previous core shroud )

inspections. '

conditions as the haus for performing flaw evaluations. .

This equated to usmg faulted condition loadings for 1 PEco's inspections of the PB-3 core shroud (completed evaluations of welds Hi-HS, and upset condition ,

loadings for evaluations of welds H6 and H7. PECo's during RFO 3R9) were performed in awsd.me with recommendations of SIL-572, Revision 1 (Ref. 26), andi LLAs of the HI-H7 welds indicated that the welds included enhanced VT-1 examinations at eight (8) cell would meet the safety margin criteria specificed in locations in each of the H1-H5 welds. PEco expanded Section XI of the ASME Code for all postulated loading conditions. b remaining ligaments of the H3 and H4 the inspection scope after discovering relevant flaw wolds were also subject to evaluation using LEFM indications at the H3 and H4 welds. W expanded methods to account for the high-neutron fluences that are scope included the following anh== we VT-1 ,

examinations:

typical at these weld elevations. The LEFM analyses of  ;

the H3 and H4 welds also indicated that the welds would have sufficient structural margin to justify operation of 100-percent examination of accessible portions of PB-2 for the cycle. the H3 and H4 welds from the ID 100-percent examination of accessible areas of b staff reviewed PECo's methodology for performing weld H4 from the OD '

flaw evaluations of the PB-2 core shroud welds, and NUREG-1544 8-12

4

. examinations of the H3 weld from the OD in areas Guidelines," Rev.1, dated April 21,1995 (Ref. 22), I 1

~

where cracking was not indicated on the ID and with the inspection scope previously approved for the PB 2 core shroud, which was inspected in the Fall of

. an examination of the H3 weld from the OD in 1994. The NRC therefore concluded that PEco's areas where cracking was indicated on the ID proposed scope for inspection of the PB-3 core shroud was acceptable for implementation during RFO 3R10.

i . examinations at six locations of the H6 weld The NRC accepted PEco's proposal for the UT inspection scope on September 25,1995 (Ref. 80).

. exammations at two locations of the respective H7 and H8 welds 8.11.5 Peach Bottom Core Shroud Repair Designs j . examination of a vertical weld between the H3 and By letter dated September 16, 1994, PECO Energy

H4 welds Company (PECO) submitted the design details of a proposed core shroud stabilizer design for PB2 and PB3

. examination of one of the mid shroud plates. (Ref. 81). The Peach Bottom core shroud repair 4

involves the installation of GE-designed tie rod i b licensee's VT-1 examinations identified a significant assemblies symmetrically around the circumference of i '(2.67 m or 105 in) crack in the H3 weld (the weld the shroud. These tie-rod assemblies were designed to j joining the top guide support ring to the upper mid- provide an alternative load carrying capability for the shroud shell). less extensive cracking was also found shroud in lieu of shroud welds H1- H7 during normal l ct the H4 weld (< 0.76 m or 30 in total). Minor operating, transient, and postulated design basis accident l i cracking was determined to exist at weld H1 and at one and seismic conditions. The Peach Bottom repair design 1 of the vertical shroud welds, was submitted as an alternative to the requirements of the American Society of Mechanical Engineers (ASME) .

s PEco performed flaw evaluations of the PB-3 shroud in Boiler and Pressure Vessel (B&PV) Code, pursuant to I

accordance with the structural margin criteria found 10 CFR 50.55a(s)(3)(i). The staff is currently in the t

GENE-523-141 1093, Rev.1 (Ref. 78). These criteria process of reviewing PECo's submittal.

conform to the structural margin criteria found in

.Section XI of the ASME Code. The evaluations of the PB 3 - core shroud indications, which included 8.12 Power Authority of the State of New York adjustments to account for crack proximities, crack l growth and NDE uncertainties, indicated that the PB-3 8.12.1 Assessment of the Response to GL 94-03 for the core shroud would meet the safety margin criteria James A. FitzPatrick Nuclear Power Plant  !

specified in Section XI of the ASME Code to justify operation during the current operating cycle (Operating The Power Authority of the State of New York  ;

Cycle 10). The staff reviewed PECo's inspection results (NYPA), the licensee for the James A. FitzPatrick  ;

and flaw evaluation methods, and concluded that they Nuclear Power Plant (FITZ), responded to GL 94-03 on were acceptable to justify continued operation of PB-3 ' August 24,1994 (Ref. 82), as supplemented with for Operating Cycle 10 (Ref. 74). responses on October 18,1994, and November 30,1994 (Refs. 83 and 84). The NRC staff considered the FITZ 8.11.4 Reinspection Scope for the Peach Bottom core shroud to be highly susceptible to IGSCC based on Unit 3 Core Shroud the following determinations:

By letter dated June 16,1995, PECo provided the NRC (1) The average reactor coolant conductivity at FITZ )

with its supplemental response to GL 94-03 (Ref. 79),

during the first five years of plant operation was This submittal provided PECo's scope for performing high in comparison to the industry norm.

UT examinations of the PB-3 core shroud during RFO 3R10, in September / October of 1995. The NRC '(2) W FITZ core shroud was fabricated from Type determined that PECo's proposed inspection scope was 304 cut and rolled plate materials, which are consistent with the guidelines of the BWRVIP "BWR considered to be more susceptible to IGSCC than Core Shroud Inspection and Flaw Evaluation forged Type 304 or Type 304L stainless steels.

I 8-13 NUREG-1544

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

. (3) Weld residual stress levels resulting from Category "C" shrouds.  % BWRVIP also fabrication of the shroud were considered to he recor -M that TVA perform comprehensive  ;

high. inspections of the BF-1, BF-2, and BF-3 core shrouds.

I b staff concluded that the BWRVIP's assessment of herefore, the staff concluded that the BWRVIP's the BF-1, BF 2 and BF-3 core shrouds was appropnate ,

ranking of the FITZ core shroud as a highly susceptible (Ref. 89). However, based on a review of the plant-Category 'C" shroud was appropriate (Ref 85). specific IGSCC susceptibility factors, the staff concluded that, while the BF1, BF-2 and BF-3 core shrouds were  ;

8.12.2 Repair of the Jannes A. FittPatrick Core Shroud likely to contain some cracking, the extent of any j IGSCC would be less than that identified at other highly l NYPA originally indicated that inrpections of the FITZ susceptible BWRs. N results of the BF-2 and BF-3 core shroud would involve 100% UT inspections of all core shroud examinations confirm this conclusion for .

accessible areas on shroud welds H1-HS, UT and Units 2 and 3.

nnhancarl VT-1 exanunations of welds H6a and H6h, and enhanced VT-1 inspections of welds H7, HB, and 8.13.2 Inspections of the Browns Ferry {

H9. On October 21,1994, NYPA informed the staff Units I,2, and 3 Core Shrouds i that it would perform a pre-emptive modification of the FITZ shroud inasaad of a comprehensive chroud in its response to GL 94-03, TVA stated that BF-1 has j mapection (Ref. 86). The shroud modification was been in an prolonged defueled condition since 1985, and  ;

designed by MPR for the purpose of providin;' an that no scheduled restart date has been scheduled for the  !

alternative load path for the reactor core in lieu of ti:e reactor. TVA indicated, however, that should a decision ~!

core shroud during normal operating, transient and be made to restart BF-1, the core shroud would be l postulated design basis accident conditions. inspected before the unit was restarted. The inspection results would then be evaluated to justify operation of ,

b licensee also informed the staff that the inspect;on the unit for the upcoming cycle. ]

scope for the FITZ core shroud would be revised to F

support inf mentation of the core shroud modification TVA stated that BF-3 has also been in a prolonged  ;

design. N revised inspection scope included, as a defueled condition. However, TVA completed an mininum, VT-1 mspections of tne welds joining those

.)

inspection of the BF-3 core shroud on July 14, 1994.

gusset plates used in the repair design to the jet pump . W results of the BF-3 shroud inspections were  ;

support plate and the reactor pressure vessel; UT presente<l to the NRC - during a meeting held on inspection of at least one vertical seam weld below the Auguac 4 1,1994, and were submitted as an enclosure to H4 weld; and inspections of the H3, H6a, and H6h the lice. Wee's GL 94-03 response. The inspection ,

wolds to gauge the extent of cracking in the shroud identified cracking at three weld locations on the BF-3 support rings, la the staff's SER dated January 5,1995, core shroud. However, TVA determined that the extent i the staff concluded that both NYPA's core shroud repair of cracking in the BF 3 core shroud was sinttv3 TVA, design and reduced core shroud inspection scope were in conjunction with GE, completed an analysis which acceptable for implementation at the plant sh (Ref. 87). demonstrated that the BF-3 core shrovi had vequate margin to justify operation of the unit. The staff reviewed the inspection results and flaw evaluation 8.13 Tennessee Valley Authonty Core Shrouds regarding the BF-3 core shroud, and found TVA's determination to be acceptable. Since both BF-1 and <

8.13.1 A==a====at of the Response to GL 94-03 fv BF-3 have been idle since 1985, TVA did not submit a Browns Ferry Nuclear Plant Units 1,2, and 3 justification for continued operation (JCO) for these units.

T- Valley Authority (TVA), the licensee for Browns Ferry Nuclear Plant Units 1, 2, and 3 TVA informed the NRC that it had scheduled inspections  !

(BF-1, BF-2 and BF-3, respectively), responded to of the BF-2 unit for the September 1994 RFO. TVA j GL 94 03 on August 24,1994 (Ref. 88). In the report completed the inspections of the BF-2 core shroud in t

'BWR Core Shroud laspection and Flaw Evaluation October 1994, and submitted the results to theNRC on 1 Guidaliaan,* .(Ref.19), the BWRVIP classified the November 18,1994 (Ref. 90). W scope of TVA's Bsowns Ferry shrouds as being highly susceptible inspection of the BF4 core shroud covered portions of a

NUREG-1544 8-14

i l

l i

\

l 1

welds HI-H7. TVA performed the UT inspections 8.15 Plants with Cateeorv *B' Core Shrouds I using 45' shear wave and 60' longitudinal wave  !

transducers. TVA also used UT creeping wave methods in late August 1994, Commonwealth Edison Co. l to determine near side surface conditions. TVA (Comed), Philadelphia Electric Co. (PECo), i i6tified cracking in or adjacent to welds H2, H3, and Pennsylvania Power & Light Co. (PP&L), and Entergy l 115. The indications, however, were minor. ne largest Operations, Inc. (EOI) submitted their responses ,

linear crack was located at weld H3 and was less than regarding the core shrouds at Lasalle Units 1 and 2 1 0.13 m (5 in) in length. ne deepest indication was (LA-1 and LA-2), Limerick Unit 1 (LIM-1), l measured to be 0.024 m (0.% in). Susquehanna Units 1 and 2 (SSES 1 and SSES-2), and  !

Grand Gulf Unit 1 (GG-1), respectively (Refs. 41,73, j

' Cracks identified during the inspections were initially 93 and 94). The staffissued its SERs regarding these compared to plant-specific screening criteria. submittals on February 16,1995, March 7,1995, -

%escreening criteria were consistent with the evaluation March 23,1995, and March 29, 1995, respectively guidelines established by the BWROG in the 'BWR (Refs. 93-%).

Core Shroud Evaluation,' GENE-523-148-il93, dated April 5,1994 (Ref.14). Thess criteria are similar to la order to assist these licensees with their submittals to those approved for use in the flaw evaluations of the staff, the BWRVIP performed a susceptibility cracking in the Brunswick Nuclear Plant Unit I core assessment of these shrouds, and concluded these shroud. All indications in the BF-2 core shroud were shrouds were not as highly susceptible to IGSCC as the determined to he below the inapection screening limits of core shrouds of Category 'C' plants. The BWRVIP 3 the evaluation guidelines. The staff therefore concluded based its assessment of these shrouds on the following I that the BF-2 reactor could safely operate for the current factors: l operating cycle without requiring a nwxlification of its core shroud (Ref. 89). (1) The construction material for fabrication of these shrouds was low carbon content Type 304L stainless steel (a more 1GSCC resistant material 8.14 Ver==* Ya Nuclear Power Cornoration than Type 304 stainless steel).

I 8.14.1 Assessment of the Response to GL 94-03 for the (2) The reactor coolant chemistry impurity levels at i Vermont Yankee Nuclear Power Station these facilities were typically maintained at lower levels than at the norm for the industry.

Vermont Yankee Nuclear Power Corporation (VYNPC), {

. the licensee for the Vermont Yankee Nuclear Power The BWRVIP concluded, however, that some potential l Station (VY), responded to GL 94-03 on existed for cracking to initiate in these shrouds based on  ;

August 17,1994 (Ref. 91). The staff issued its SER the amount of time that these plants had operated at '

regarding VYNPC's response to GL 94-03 on power (> 8 years of power operation). As a result, the January 5,1995 (Ref. 92) BWRVIP categorized these shrouds as Category 'B' shrouds and recommended that the licensees owning VYNPC inspected the VY core shroud during the these facilities inspect the circumferential welds Spring 1995 RFO, which commenced on associated with the top guide support ring, core support March 18,1995. (Section 5.5 of this report discusses plate ring and mid-shroud shell at the next available the scope and results of the VY core shroud RFO. (Table 6.2-1 summarizes the BWRVIP's examinations.) ne staff reviewed the VY core shroud susceptibility rankings and inspection recommendations.)

inspection results and fisw evaluations in April 1995, and i.cs its e SER regarding these submittals on May 5, Since July 1994, the licensees owning these Category 1995 (Ref. i3), in that SER, the staff concluded that "B" shrouds have performed limited inspections of the '

the VY core shroud had sufficient structural margin to LA-1, LA 2, LIM-1, SSES-1, and GG-1 shrouds, justify one additional cycle of operation for the plant. PP&L has scheduled inspections of the SSES-2 core 8-15 NUREG-1544

shroi.d for the Fall 1995 (Ref. 93). By letter dated The BWRVIP based its assessments of these shrouds on May 22,1995, PP&L submi;ted its scope for performing thef awing factors-  !

UT snepections of the SSES-2 core shroud (Ref. 99).

i The staff determined that PP&L's inspection scope for (1) The construction material for fabrication of these SSE!i-2 was consistent with the BWRVIP's inspection shrouds was low carbon content Type 304L criteria for Category "B" shrouds. & staff accepted stainless steel (a more IGSCC resistant material of l the inspection scope for the SSES-2 core shroud on construction).

~

August 7,1995 (Ref.100).

(2) The reactor coolant water chemistries at these Of the inspections performed on these core shrouds to plants, when averaged over the first five years of date, only the inspections of the SSES-1 core shroud operation, were typically better than the norm for have resulted in the identification of any IGSCC-related the industry, craci ing in excess of 2.54 m (100 in). PP&L submitted the results of the SSES-1 core shroud UT examinations (3) These reactors had been operated for only a on April 21,'1995 and May 25,1995 (Refs.' 101 and limited amount of time at power, in comparison to  !

102). Of particular note were the reports of indications other BiVRs in the industry (< 8 years total time i located in the SSES-1 H 4 and H-5 shroud welds. % at power).

flaw lengths at these welds totaled 4.75 m (IG7 in) and i

4.80 m (189 in), respectively. 'Ilis cracking is The BWRVIP therefore grouped the core shrouds of  ;

significant in that it was the first report of moderate these plants as Category "A" type shrouds, and ,

cracking in Category "B" core shrc9ds to date. PP&L suggested that licensees owning these facilities could ,

provided its flaw evaluation of the SES 1 core shroud defer examinations of their core shrouds until the first in its submittals to the NRC (Refs.101 and 102). The RFO commencing after their plants had surpassed 8 '

reauks of PP&L's flaw evaluation indicate that the years of time at power. Upon reaching that point, the j SSES-1 core shroud will meet the ASME Code, BWRVIP recommended that the licensees for these i Section XI safety margins for the current cycle. On plants conduct examinations of these core shroud in '

May 3,1995, the NRC accepted the results of PP&L's accordance with the BWRVIP's recommendst ons for flaw evaluation and approved the restart of SSES-1 Category "B" shrouds (Refer to Section 7.2.15 and

{

(Ref.103). Table 6.2-1 of this report.). '

The NRC reviewed the BWRVIP's assessment of these .;

8.16 Plants with Catenorv "A" Core Shrouds shrouds, and concluded that the sesessments were appropriate based on the factors cited in the BWRVIP In August 1995, the Philadelphia Electric Co. (PECo), {

assessment. '!he staff therefore concluded that the

' Public Services Electric and Gas Company (PSE&G), licensees owning these plants could safely defer l Niagara Mohawk Power Corp. (NMPC), Entergy inspection of their shrouds until they had surpassed 8  ;

Operations, Inc. - (EOI), Detroit Edison Company years at power (Refs. 111 118). i

_(Deco) Illinois Power Co. (IPC), Centerior Energy, i Inc. (CEI) and Washington Public Power Supply System To date, PECo (the licensee for LIM-2), PSE&G (the (WPPSS) submitted their a ,~..a regarding the core licensee for HC-1), IPC (the licensee for the CPS), and {

shrouds at Limerick Unit 2 (LIM 2), Hope Creek Station EOI (the licensee for the RVR 1), have deferred  !

Unit 1 (HC 1), Nine Mile Point Unit 2 (NMP-2), examinating their core shrouds at this time. Deco, River Bend Unit I (RVR-1), Fermi Unit 2 (FRM-2), l NMPC, and WPPSS (the licensees for FRM 2, NMP-2, Clinton Power Station (CPS), Perry Nuclear Power Plant WNP-2, respectively) performed limited examinations of (PRY) and Washington Nuclear Plant Unit 2 (WNP-2), their core shrouds during the last RFOs for their nuclear respectively (Refs. 73, 104 - 110). In order to assist facilities. No IOSCC-related crackmg indications have these licensees in their submittals to the staff, the been identified as a result of the limited examinations BWRVIP performed a susceptibility maamment of these performed by these license s.  ;

akrouds, and concluded that, of all core shrouds in the j industry, these shrouds were least susceptible to IGSCC. l NUREG-1544 8-16 I

l

9 IGSCC IN OTIIER BWR INTERNAL COMPONENTS 1

9.1 Core Plate and Top Guide Cracking core plate to occur, udlibiting control rod insertion. In this event, SLC would be required 12 November 1994, IGSCC was reported in the top for reactor shutdown. However, simultaneous guide and core plate rings of a foreign BWR. On failure of all of the core plate hold-down bolts is Novemeber 22,1994, GE issued RICSIL 071 to inform highly improbable.

the U.S. nuclear industry of this cracking (Ref. I19). In RICSIL 071, GE concluded that, while the design of the (3) Lateral motions is also prevented in plants with foreign BWR was not a GE design, there were enough hold-down bolts and four horizontal aligner pin j similarities between the design of the foreign BWR and assemblies, j those of U.S. BWRs to warrant an investigation to determine whether domeatic BWR top guides and core (4) Vertical displacement of the core plate during plates would similarly be susceptible to IGSCC. GE, design basis loading is limited to 0.013 m however, did not provide any recommendations in (0.5 in) due to the clearance between the core RICSIL O71 in regard to performing inspections of BWR support plate and the fuel support structures.

top guides and core plates. Vertical displacements of this magnitude were evaluated by the staff during the core shroud In December 1994, the NRC requested that the assessments and were not found to inhibit control BWRVIP provide them with the details of top guide and rod insertion.

core plate configurations in domestic BWR designs, and assess the safety significance of the top guide and core (5) Vertical displacement of the top guide during a plate cracking at U.S. BWR facilities. A preliminary LOCA with postulated through-wall cracking of response from the BWRVIP was received on the top guide ring weld is bounded by analyses December 23, 1994, and a revised response was performed during the core shroud assessments.

J received on January 3,1995 (Refs.120 and 121) GE subsequently issued SIL 588 (February 17,199D to (6) A variety of designs exist for lateral and vertical update their position on safety significance and pn& restraint of top guides. Certain configurations are specific recommendations for inspections of BWR top more susceptible than others to lateral guides and core plates (Refs.122). The NRC reviewed displacements under design-basis loading.

the information submitted by the BWRVIP and GE Specifically, for top guide designs which do not regarding top guide and core plate cracking. Upon incorporate wedges or reinforcement blocks, completion of thier review, the staff came to the failure of the pin / aligner supports due to IGSCC following conclusions (Ref.123): could result in lateral displacement of the top guide during seismic loading. The lateral (1) With regard to IGSCC susceptibility, it is displacement could inhibit control rod insertion reasonable to expect that U.S. BWRs (BWR-2 and SLC would be required for safe shutdown.

through BWR-5) operating with conditions similar However, staff evaluations performed for GL 94-to those found at Wuergassen (13 years operating 03 have indicated that full control rod insertion time, moderate conductivity water chemistry) may would likely occur under these conditions.

experience cracking in the top guide and core plate rim ring welds. (7) Due to the potential consequences of the lateral displacement of top guide assemblies which do not (2) All U.S. BWRs have core plates with 36 to 70 incorporate wedges, SIL 588 recommends an hold-down bolts. With the hold-down bolts intact, enhanced VT-1 inspection of the members which core plate ring cracking has an insignificant impact provide the load path between the aligmnent pins, on core plate displacements under design basis the top guide and the shroud during the next  !

loading. If complete separation of all hold-down scheduled refueling outage. The staff agrees with bolts is postulated in conjunction with a seismic this recommendation. For the core shroud bolts, event, the potential exists for lateral motion of the the SIL recommends only an inspection to confirm 9-1 NUREG-1544

l that the bolting is in place. While this is most well as the assumed blowdown loads (Ref.125). As a i likely sufficient based on the redundancy of the result IGSCC of JPHDB assemblies has the potential to j structure, the staffi--+- M that consideration reduce anfety margins during postulated LOCAs. l be given to a enore compri.hensive inspection of a Failure of a JPHDB assemblies can also result in loose  ;

limited sampling of the core plate bolts. 'Ihis parts, although events to date have not resulted in any -]

reco==aarlation has been communicated to the damage to safety-related systems or equipment.  ;

BWRVIP, and the staff espects that their revised However, failure of a JPHDB can be detected during -

report on top guide / core plate cracking will power operation, thereby assuring prompt corrective address this issue. action. BWR licensees are performing inspections of l their JPHDBs and/or core flow balance tests, as required  ;

(8) With the SIL No. 588 rocc---* inspections, by Technical Specifications, to confirm the operability of '

the NRC staff concurs with the BWRVIP and GE their jet pumps.

a-= ants that the potential cracking of the top guide and core plate rings does not have a In June 1980, GE issued a SIL No. 330, " Jet Pump significant impact on safety (Ref.123). Beam Cracks" to highlight the problem of JPHDB cracking (Ref.126). As a result of the inspections  !

performed at the request of the NRC, several other  ;

9.2 Jet Pump Hold-Down Beams licensees reported cracking of JPHDBs at their facilities. t These plants include DR-2, MS-1, PB-3, PNPS, QC-1  ;

Jet pump hold-down beams (JPHDBs) provide lateral and VY. IEB 80-07 was closed out with the issuance of .

support for the jet purnp ==a=hlies at the rams head NUREG/CR-3052 (November 1984), which summarized i diffuser elevations. In February 1980, a JPHDB failed the findings and actions taken to resolve the issue of at Dresden 3 (DR-3), resulting in disassembly of one of JPHDB cracking in BWRs (Ref.127). l the plant's jet pump assemblies. As a result of the jet  ;

pump flow anomalies, Commonwealth Edison (Comed, On September 13, 1993, Entergy Operations, '

the lima =a for the Dresden, Quad City, and Lasalle Incorporated (EDI), reported the occurrence of cracking ,

BWRa) co==aa-I an orderly shutdown of the DR-3 in a JPHDB at the Grand Gulf Nuclear Power Station, unit. Subsequent visual and ultrasonic inspections, Unit I (GG-1), during the 1993 RFO for the plant. EOI -  !

conducted at the direction of GE, disclosed that hold- opted to replace all JPHDBs at the plant during the '

down beams on other jet pumps at DR-3, Quad Cities RFO. The JPHDB failure at GG-1 is unlike the earlier Unit 2 (QC-2, March 15-16,1980) and Pilgrim (PNPS, JPHDB failure at DR-3, in that the cracking at GG-1 March 28,1980) contamed cracks in the ligament zone occurred in the transition area between the main body of at the center of the beams. Investigations determined the beam and the beam end, while the failure at DR-3 that these cracks were caused by IGSCC, which in the involved IGSCC cracks initiated at the bolt hole in the case of the JPHDS failures progressed very slowly over center portion of the JPHDB. l a penod of years. j EOI expanded the inspection scope to other JPHDBs, The NRC issued laspection and Enforcement Bulletin and determined that a second JPHDB was degraded in (IEB) 80-07 'BWR Jet Pump Assembly Failure' in the center bolt-hole region (in the same portion of the April 1980 to inform the industry of the JPHDB JPHDB as the failure at DR-3). The JPHDB failures at cracking at DR-3 (Ref.124). In IEB 804)7, the NRC GG-1 are the first JPHDR failures at a GE-designed requested that lica=== owning BWR-3 and BWR-4 BWR-6 facility. Grand Gulf decided to replace all of facilities inspect their JPHDB assemblies and begin their beams during their ongoing refueling outage. On operability surveillances to justify further operation of December 17, 1993, the NRC issued IN 93101, " Jet their units. Pump Hold-down Beam Failure,' to inform the industry '

of the cracking discovered at GG-1 (Ref.125). H Failure' of a JPHDB asseinbly and subsequent sta===a=Wy of a jet pump could potentially result in an During UT inspection of the hold-down beams in increased flow area through thejet pump and lower the October 1993, the Illinois Power Company (IPC), the floodmg elevation of the core during postulated LOCAs licensee for the Clinton Nuclear Station (CNS, a BWR-6

- (Ref. 125). Such effects could adversely impact the design) determined that one of the beams (No. 7) had water level in the core during a postulated LOCA.' as crack indications around the center hole region.

NURBG-1544 9-2

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

i a

IPC replaced the cracked component before returning the Dree concerns have been identified with regard to a unit to service. In addition, on November 22,199't, postulated failure of an access hole cover due to IOSCC.

i Pennsylvania Power and Light Company (PP&L) i l notified the NRC that they would be rep iecing all of the 1. loose parts - In the event of complete failure of i jet pump beams at the Susquehanna eam Electric an AHC weld during normal operation, the

! Station Unit 1 (SSES 1). This action w being taken slightly higher bottom head area pressure would cfter GE informed the licensee that tv ' could not lift the cover out of its recess. It would most l adequately predict, at this time, the crack etistion and likely fall to one side, but the potential exists for j crack growth rate of the beams, given the new failure it to be swept into the re-circulation pump suction l 2 mode of the Grand Gulf beam. De old method relied line and cause severe pump damage. l

, on predictable crack growth rates of IGSCC and cracks

! appearing at the center bolt hole region. 2. Core flow bypass (normal operation) - less of (

one or both cover plates would allow some

, GE has recommended that licenseca replace their recirculation system flow to bypass the core, from

! JPHDBs at the earliest opportunity if the JPHDBs are of the jet pump discharge through the open hole to j s design similar to that used at GG-1, and if the JPRDBs the recirculation pump suction. his flow ,

1

- have a history of being in service for more than 8 years transient would be readily detectable and would i j as of the next RFO. Although there are no requirements require reactor shutdown.

governing inspections of JPHDBs, licensees have been l mapocting their JPHDBs at their own initiative. 3. Core flow bypass (LOCA) - If an AHC weld l were to fail as a result of a RLB, the bypass path  !

! would prevent the emergency core cooling

! 9.3 Access Hole Covers system from reflooding the core to the j

! two-thirds level. However, the core spray system  !

BWRs are designed with two access hole covers (AHCs) would be capable of maintaining adequate core i in the shroud support plates. These AHCs are located cooling. I at the bottom of the annulus region between the reactor j

, vessel wall and the core shroud,180* apart from each GE issued SIL 462, including Supplement 1 j

{ other, where they provided access to the lower plenum Supplement 2, Supplement 2 Rev.1, and Supplement 3, tress during construction phases. AHC are mainly provided the industry with information regarding AHC

', fabricated from Alloy 600, and were welded to the cracking (Refs. 128-132). GE recommended that BWR l shroud support plate with Alloy 82 or 182 weld material licensees inspect the AHC welds at the next available i before initial startup of the plants. Like Type 304 and RFO if the AHC weld areas had not been previously

, 316 stainless steels, these alloys are known to be inspected. GE also recn =l~I that licensees who had

susceptible to IGSCC. examined their AHC welds review the inspection results.

i Furtherinore, GE recommended that licensees perform

} On January 21,1988, PECo reported the occurrence of suitable repairs of the flaws if IGSCC is detected

aignificant circumferential cracking in the welds joining in AHC welds. In addition, the NRC also issued

, the AHCs to the shroud support plate of the PB-3 IN 88-03, " Cracks in Shroud Support Access Hole i nuclear plant. PEco discovered these cracks using a Cover Welds" (Ref.133), and IN 92 57, " Radial remotely operated ultrasonic inspection method. PECo's Cracking of Shroud Support Access Hole Cover Welds" inspection results indicated that the cracks had initiated (Ref.134), to inform the industry of the event at PB-3.

as a result of vertical crevices at the welds, and had Although there are no requirements governing propagated along the weld fusion lines. Other cases of inspections of AHCs, licensees have been inspecting AHC weld cracking were reported as a result of their AHCs at their own imtiative.

inspections performed at the QC-2 nuclear plant.

3 NUREG-1544

10 CONCLUSIONS AND FUTURE ACTIONS The NRC staff has revi-wed the licensee-specific As of early September 1994, the NRC staff received all responses to GL 94-03. In a3 cases, the staff has of the BWR licensee submittals in response to GL 94-03.

concluded that the BWR licensee have provided The staff has completed its evaluations of the licensee sufficient evidence to support continued operation of responses to GL 94-03, and has transmitted the SERs to their BWR units until the refueling outages during which the appropriate BWR licensees. For all cases, the staff shroud inspections or repairs have been scheduled. The concluded that BWR licensee's have provided sufficient results of inspections performed by the industry indicate justification to operate their facilities until core shroud that IGSCC can occur in BWR core shrouds fabricated inspections or repairs could be implemented. The staff from Type 304 stainless steel rolled plate materials. based its conclusions on the following factors:

Prior to issuance of GL 94-03, the NRC analyzed the (1) To date, no 3600 through-wall core shroud results of inspections and flaw evaluations performed at cracking has been observed in any U.S. BWR that Brunswick Unit 1. Dresden Unit 3, and Quad Cities has performed a shroud inspection.

Unit I because of the severity of the flaw indications at these plants. In light of the extent of cracking observed (2) All analyses performed by U.S. licensees to date et these plants, the staff evaluated potential safety indicate that, even if cracking did exist in a concerns associated with the possibility of a 360* particular BWR core shroud, sufficient ligaments circumferential separation of the shroud following a would remain in the shroud such that structural 1 postulated LOCA. He staff's evaluation considered the integrity of the shroud would be ensured for the potential for separation of the shroud during postulated remainder of the plant's operating cycle.

accidents to either prevent full insertion of the control rods, or open a gap large enough to preclude the ECCS (3) No U.S. BWR has exhibited any of the symptoms from fulfilling their intended safety functions. He (power-to-flow mismatch) that would indicate bounding case accident scenarios are the MSLB and the leakage through a 3600 through sal; mi ~ud crack.

RLB. Of these postulated accidents, the MSLB is the more serious event associated with cracks in the upper (4) MSLBs or RLBs are both considered to be low shroud welds (e.g., H2, H3), and the RLB is the more frequency events.

serious event associated with cracks in the lower elevations of the core shroud. (5) Only short durations remained until core shroud inspections or repairs would be implemented by the In consideration of the consequences of a 360* through- individual BWR licensees.  ;

wall failure of the shroud coincident with a LOCA, the NRC has conservatively estimated the risk contribution Since January 1994, the staff has reviewed and accepted from shroud cracking and determined that it does not the design modifications submitted in regard to repair of pose a high degree of risk at this time. However, the the core shrouds at Brunswick Units I and 2, Hatch NRC has also determined that ASME Code structural Units 1 & 2, FitzPatrick, Oyster Creek, Quad Cities margins could potentially be exceeded if the cracks were Units 1 & 2, Nine Mile Point Unit 1, and Pilgrim. The sufficiently deep and were to continue propagating staff is currently reviewing the design modifications through the shroud during normal operating, transient, submitted by the Commonwealth Edison Company or accident conditions, which could result in a loss of a (Comed) and Philadelphia Electric Company (PEco) in 1;yer of the defense-in-depth strategy. H erefore, in regard to repair of the core shrouds at Dresden Units 2 order to verify compliance with the inservice inspection and 3, and Peach Bottom Units 2 and 3. Additional requirements of 10 CFR 50.55a and to ensure that the design modifications will be submitted if licensees risk nasociated with core shroud cracking remains low, determine that their shrouds are highly susceptible to the staff has concluded that it was appropriate for BWR IGSCC, or if inspection results indicate that large-scale licensees to implement timely inspections and/or repairs cracking of circumferential shroud welds has occurred.

of their core shrouds. Rese repairs or modifications are designed to ensure the structural integrity of the core shrouds in the long term.

10-1 NUREG-1544

l I

l i

and are being reviewed by the NRC staff on a case-by-case basis. 'Ibe staff will continue to assess the scopes that have yet to be subunitted by lic- concerning

'-- ;- A== or reinspections of their core shrouds. The

, staff will also continue to assess core shroud inspection l results and any appropnete core shroud repair designs on I a case-by case basis. The staff will issue separate SERs regarding the acceptability of core shroud inspection results and core shroud repair designs.

l l The BWROG has formed an independent organization, the BWRVIP, for the pugmee of providing recommaantaesans and guidelines in regard to inspections, l evaluations, and repairs of BWR internal components.

l The BWRVIP has submitted its initial guidelines ,

regarding inspections, evaluations, and repairs of BWR l core shrouds, and its guidelines regarding qualification l cf NDE techniques, to the staff. These guidelines have i been reviewed and accepted by the staff. The BWRVIP  !

has also submitted its revised criteria regarding qualifications of NDE techniques for inspection of BWR internal components, standardized repair submittal formats, and generic anfety assessment regarding BWR internal components. These documents are currently under review by the staff. 'Ihe BWRVIP has committed to submit in 1996 its recommended guidelines for performing reinspections of BWR core shrouds, and its reca==aadant guidelines for performing augmented inspections of BWR core shroud repair assemblies. The staff will review these documents following their receipt.

The staff will continue to request timely submittals and responses by the BWRVIP Technical Subcommittees and individuallicensees to meet the established schedules for plant startups. The staff has interacted with the BWRVIP and individuallicensees on a request basis to achieve this goal. The staff will continue to interact with the industry in the future to encourage th'e m in their efforts to address IGSCC and other forms of age-related degradation in BWR internal components.

1 l

NUREG-1544 10-2 l

11 REFERENCES 1 Memorandum from James M. Taylor, Executive Director of Operations to the NRC dated January 4,1994, " Emerging Technical Issue: Cracking of BWR Vessel Internals." i 2 NRC Generic Letter 9443, dated July 25,1994, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors." )

3 General Electric Company RICSIL 054, dated October 3,1994, " Core Support Shroud Crack Indications."

4 General Electric Company RICSIL 054 Rev.1, dated July 21,1993, " Core Shroud Cracks."

5 NRC Informaison Notice 93-79, dated September 30,1993, " Core Shroud Cracking at Beltline Region 1 Welds in Boiling Water Reactors."

l 6 Letter from the NRC to R. A. Anderson, Vice President, Brunswick Nuclear Project, Carolina Power and Light Company, dated January 14,1994, " Evaluation and Repair of the Core Shroud Cracks, Brunswick Sacam Electric Plant, Unit 1 (TAC No. M87270).*

7 NRC Informosum Notice 94-42, dated June 7,1994, " Cracking in the Lower Region of the Core Shroud in Boiling.Waser Reactors."

8 NRC Informaamm Notice 94-42, Supplement 1, dated July 19,1994, " Cracking in the Imer Region of the Core Shrumi in Boilms-Water Reactors."

9 letter from the NRC to D. L. Ferrar, Manager, Nuclear Regulatory Services, Commonwealth Edison Company. W Jii) 21,1994, " Evaluation of Core Shroud Cracking at Dresden, Unit 3, and Quad Cities, Unis I (TAC Noa. M89871 and M89493)."

10 letter from R. W. Kessen, Vice President and Director of Technical Functions, GPU Nuclear Corporation, so the NRC, dated October 25,1994, " Oyster Creek Nuclear Generation Station Core Shroud Repaar - Design Report," Rev. O.

11 letter from the NRC to J. J. Barton, Vice President and Director, GPU Nuclear Corporation, dated  ;

November 25,1994, " Safety Evaluation Regarding the Oyster Creek Core Shroud Repair -

(TAC No. M90104)."

12 1 =**ar from Vermont Yankee Nuclear Power Corporation to the NRC forwarding the Memorandum from S. R. Miller. Yankee Atomic Electric Company to R. D. Pagodin, Vermont Yankee Nuclear  !

Power Corporation, dated April 20,1995, " Vermont Yankee Core Shroud Inspection and Flaw Evaluations."

13 letter from the NRC to D. A. Reid, Vice President of Operations, Vermont Yankee Nuclear Power Corporation, dated April 27,1995, " Core Shroud Inspection and Flaw Evaluation, Vermont Yankee i Nuclear Power Station (TAC No. M92050)." l 14 Letter fmm L. A. England, Chairman, BWROG, to the NRC, dated April 5,1994, BWR Core Shroud Evaluation (GENE 523148-1193)." l l

11-1 NUREG-1544  !

l l

)

15 BWROG/GE letter to the NRC, dated July 13,1994, 'BWR Shroud Cracking Generic Safety Assessment (GENE-523-A107-0794)."

16 letter from R. A. Pinelli, Chairman BWROG, to the NRC, dated Aug. 5,1994, "BWR Shroud Cracking Generic Safety A=-mt," Rev, I.

17 letter from R. A. Pinelli, Chairman BWROG, to the NRC, dated Aug. 5,1994, " Responses to NRC Questions on Core Shroud and Reactor Internals."

18 Letter from the NRC to S. LaBruns, Chairman, BWROG Executive Oversight Committee, dated May i 12,1994, " Generic Questions Regarding Core Shroud Cracking.'

19 letter f-m C. Terry, Executive Chairman, Ammaamment Subcommittee, BWR Vessels and Internals  ;

Projet o the NRC, dated Sept. 2,1994, "BWR Core Shroud Inspection and Evaluation Guidelines.' l l

20 Letter from Brian W. Sheron, Director- Division of Engineering, Office of Nuclear Reactor ]

Regulation, NRC, to J. T. Beckham, Chairman, BWRVIP, dated hp*ah 13,1994, " Evaluation of BWR-VIP Report Entitled 'BWR Core Shroud Inspection and Evaluation Guidelines,'

GENE-523-113-0894, September 2,1994."

21 Letter from the NRC to J. T. Beckham, Chairman, BWRVIP, dated Dec. 28,1994, " Evaluation of  !

'BWR Shmud Cracking Generic Safety Assessment, Revision 1,' GENE-523 A107P-0794, August 5, 1994 and 'BWR Core Shroud Inspection and Evaluation Guidelines,' GENE-523-ll3-0894, September 2,1994.* i J

22 Letter from C.D. Terry, Executive Chairman - Assessment Subcommittee, BWR Vessel and Internals ,

Project, to the NRC, dated April 21,1995, 'BWR Core Shroud Inspection and Evaluation Guidelines, j Revision 1." l 23 letter from the NRC to J. T. Beckham, Chairman - BWRVIP, dated June 16,1995, ' Evaluation of *

'BWR Core Shroud Inspection and Evaluation Guidelines,' GENE-523-il3-0894, Revision 1, dated March 1995, and 'BWRVIP Core Shroud NDE Uncertainty and Procedure Standard,' ,

dated November 22,1994.  ;

24 Letter from B. McCleod, Technical Chairman, BWRVIP Repair Technical Subcommittee, to the NRC, dated Aug. 18,1994, "BWR Core Shroud Repair Design Criteria.' , .

25 letter from the NRC to B. McCleod, Technical Chairman, BWRVIP Repair Technical Subcommittee, dated Sept. 29,1994, " Safety Evaluation on Boiling Water Reactor (BWR) Core Shroud Repair Design Criteria."

26 GE SIL 0572, Rev.1, dated October 4,1993, " Core Shroud Cracks."

27 GE RICSIL 068, Rev.1, dated April 14,1994, " Update on Core Shroud Cracking."

28 GE RICSIL 068, Rev. 2., dated May 6,1994, " Update on Core Shroud Cracking.'

r 29 Letter from L. J. Olivier, Vice President of Nuclear Operations, Boston Edison Company (BECo), to J the NRC, dated August 27,1994, " Response to Generic letter 94-03, Integranular Stress Corrosion Cracking of Core Shrouds."

NUREG-1544 11 2 5

30 latter from E. T. Boulette, Ph.D., Senior Vice President - Nuclear, Boston Edison Company (BECo), to the NRC, dated October 13,1994, " Pilgrim Shroud Analysis."

31 Letter from L. J. Olivier, Vice President of Nuclear Operations, Boston Edison Company (BECo), to

- the NRC, dated Oct. 28,1994, "NRC Request for AdditionalInformation, Pilgrim Shroud Analysis."

(Response to RAI)

32 NRC letter to E. T. Boulette, Ph.D., Senior Vice President - Nuclear, Boston Edison Company, dated 3 November 28,1994, " Safety Evaluation Report for Pilgrim Nuclear Station Regarding Generic Letter 94-03: 'Integranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors' (TAC No, M90108)."

1 33 Letter from E. T. Boulette, Ph.D., Senior Vice President - Nuclear, Boston Edison Company (BECo), to the NRC, dated January 161995, " Pilgrim Nuclear Power Station Core Shroud Stabilizer i Design."

34 latter from E. T. Boulette, Ph.D., Senior Vice President - Nuclear, Boston Edison Company.

4 (BECo), to the NRC, dated February 24,1995, "PNPS Response to the NRC Staff Request for

, AdditionalInformation Concerning the Proposed Repair of the Pilgrim Core Shroud."

[ 35 Letter from E. T. Boulette, Ph.D., Senior Vice President - Nuclear, Boston Edison Company (BEco), to the NRC, dated March 21,1995, "PNPS Response to the NRC Staff Request for

AdditionalInformation Concerning the Proposed Repair of the Pilgrim Core Shroud."

i E 36 Letter from E. T. Boulette, Ph.D., Senior Vice President - Nuclear, Boston Edison Company

, .(BECo), to the NRC, dated April 14,1995, " Additional Information Concerning Our Planned Modification of the Pilgrim Core Shroud."

37 I. meter from the NRC to E. T. Boulette, Ph.D., Senior Vice President - Nuclear, Boston Edison

Company, dated May 12,1995, " Safety Evaluation Regarding the Pilgrim Nuclear Power Station Core

{

l Shroud Repair (TAC No. M91305)."

38 Letter f om R. A. Anderson, Vice President, Carolina Power & Light Company, to the NRC, dated August 24,1995, " Brunswick Steam Electric Plant, Unit Nos. I and 2, Docket Nos. 50-325 and 50-324/ License Nos. DPR-71 and DPR-62 NRC Generic latter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors." (Response to GL 94-03)

J

39 Latter from R. A. Anderson, Vice President, Carolina Power & Light Company, to the NRC,' dated April 28,1995, " Brunswick Steam Electric Plant, Unit No.1, Docket Nos. 50-325/ License Nos. DPR-

. 71, NRC Generic Letter 94-03, 'Intergranular Stress Corrosion Crackmg of Core Shrouds in Boiling i

' Water Reactors" (re-inspection results).

40 Letter from the NRC to R. A. Anderson, Vice President - Carolina Power and Light Company, dated l January 3,1995, " Generic Letter 94-03, 'Intergranular Stress Corrosion Crackmg of Core Shrouds in Boiling Water Reactors,' Brunswick Steam Electric Plant, Units I and 2 (TAC Nos. M90084 and M90085."

41 . Letter from G. G. Benes, Nuclear Licensing Administrator, Commonwealth Edison, to the NRC, dated l August 23,1994, "Dresden Nuclear Power Station Units 2 and 3, Quad Cities Nuclear Power Station Units 1 and 2, Lasalle County Nuclear Power Station Units 1 and 2 Commonwealth Edison (Comed) ,

Response to NRC Generic Letter (GL) 94-03, 'lategranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors,' NRC Docket Nos. 50-237/249,50-254/249,50-373/374."

11-3 NUREG-1544

i 42 Letter from M. D. Lyster, Site Vice-Presulent - Dresden Station, to Mr. William T. Russell, Director Office of Nuclear Reactor Regulation, NRC, dated June 13,1995. " Analytical Evaluatum of Cracking identified at Dresden Nuclear Power Station Unit 3. NRC Docket No. 50-249."

43 letter from R. J. Walsh, Core Shroud Project Manager - Quad Cities Station, to Mr. William T.

Russell, Director Office of Nuclear Reactor Regulation, NRC, dated June 13,1995. " Analytical Evaluation of Cracking Identified at Quad Cities Nuclear Power Station Unit 1, NRC Docket No. 50-254."

44 letter from the NRC to D. L Farrar, Manager - Nuclear Regulatory Services, Commonwealth Edison Company, dated January 31,1995, " Resolution of Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors,' at Dresden Units 2 and 3, and Quad Cities, Units I and 2 (TAC Nos. M90088, M90089, M90109 and M90lI0)."

45 letter from J. L. Schrage, Nuclear Licensing Administrator, Commonwealth Edison, to the NRC, dated January 16,1995, " Quad Cities Nuclear Station Units 1 and 2, Core Shroud Modification Documents and Urut 2 Core Shroud Inspection Plan, NRC Docket Nos. 50-254 and 50-265."

46 letter from J. L. Schtsge, Nuclear Licensing Administrator, Commonwealth Edison, to the NRC, dated May 24,1995

• Design Documents for the Dresden Station Core Shroud Repair, NRC Docket No.s 50-237 and 50-249.*

47 Letter from the NRC to D. L. Farrar, Manager - Nuclear Regulatory Services, Commonwealth

. Edinos Company, dated June 8,1995, " Quad Cities Nuclear Power Station, Units I and 2, Safety Evaluatum Regardag Core Shroud Repair (TAC Nos. M91301 and M91302)."

48 Letter from R. W. Keaten. Vice President and Director of Technical Functions, Genen1 Public Utilities  :

(GPU), to the NRC, deled Aug. 24,1994, " Oyster Creek Nuclear Generating Station (OCNGS), 1 Docket No. 50 219. NRC Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in Brnleg Water Reactors.'" (Response to Generic Letter 94-03) 49; Letter from the NRC to J. J. Barton, Vice President and Director, GPU Nuclear Corp., dated Novemher 25.1994,

• Safety Evaluation Regarding the Oyster Creek Core Shroud Repair (TAC No.

M90104).*

50 Letter from J. T. Beckham, Vice Preeident - Plant Hatch, Georgia Power Company (GPC), to the NRC, dated August 24,1994, "Edwin I. Hatch Nuclear Plant, BWR Core Shroud Cracking, Response to Generic letter 94 03.* ,

51 Letter from the NRC to J. T. Beckham, Vice President - Plant Hatch, Georgia Power Company <

(GCP), dated Feb. 23, 1995, SER regarding " Generic Letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs - Edwin I. Hatch Nuclear Plant, Units I and 2 (TAC Nos. M90094 and M99095).*

52 letter from J. T. Beckham, Vice President - Plant Hatch, Georgia Power Company (GPC), to the NRC, dated September 2,1994, "Edwin I. Hatch Nuclear Platt Unit 1, Installation of Core Shroud Stabilizers for Core Shroud Stabiliser Design Submittal."

53 Letter from the NRC to J. T. Beckham, Vice President - Plant Hatch, Georgia Power Company 1 (GCP), dated E;W- e 30,1994, " Safety Evaluation Report for Core Shroud Stabilizer Design, Edwm L Hatch Nuclear Plant, Unit 1 (TAC No. M90270)."'

NUREG-1544 11-4

., .- . . .- -~. - - - - - - . - . - - .. .-

n 4

i 54 letter from J. T. Beckham, Vice President - Plant Hatch, Georgia Power Company (GPC), to the NRC, dated July 3,1995, "Edwin 1. Hatch Nuclear Plant - Unit 2, Core Shroud Stabilizer Design Submittal."

1 55 Letter from the NRC to J. T. Beckham, Vice President - Plant Hatch, Georgia Power Company, j dated September 25,1995, " Safety Evaluation for Core Shroud Stabilizer Design, Edwin I. Hatch Nuclear Plant, Unit 2 (TAC No. M92783)."

56 Letter from J. F. Franz, Vice President - Nuclear, IES Utilities Inc., to the NRC, dated August 24, ,

l 1994, "Duane Arnold Energy Center, Docket No. 50-33?, Op. License No. DPR-49, Response to NRC i Generic letter 94-03, lategranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors, dated July 25, 1994."

l

57 NRC letter to Mr. Lee Liu, Chairman of the Board and Chief Executive Officer, IES Utilities Inc.,

dated March 1,1995, SER regarding the "Duane Arnold Energy Center - Response to Generic Letter (GL) 94 03, 'Intergranular Stress Corrosion Cracking of Core Shroud in BWRs (TAC No. M90090)." ,

58 Letter from G. R. Horn, Vice President - Nuclear, Nebraska Public Power District (NPPD), to the NRC, dated Aug. 26,1994, " Response to Generic Letter 94 Core Shroud Cracking, Cooper Nuclear Station Docket No. 50-298, DPR-46."

$9 letter from the NRC to G. R. Horn, Vice President - Nuclear, Nebraska Public Power District, dated April 12,1995, SER regarding " Generic letter (GL) 94-03, 'Intergranular Stress Corrosion Cracking ,

of Core Shrouds in BWRs' - Cooper Nuclear Station (TAC No. M90087)."

60 letter from G. R. Horn, Vice Presulent - Nuclear, Nebraska Public Power District, to the NRC, dated July 14,1995, " Core Shroud Inspection Plan, Cooper Nuclear Station, NRC Docket 50-298, 3 DPR-46."

61 Letter from the NRC to G. R. Horn, Vice President - Nuclear, Nebraska Public Power District, dated ,

y' '+- 20,1995, " Cooper Nuclear Station - Generic Letter 94-03, Core Shroud Inspection Plan (TAC No. M93007)."

62 letter from C. D. Terry, Vice President - Nuclear Engineering, Niagara Mohawk Power Corporation, to the NRC, dated August 23,1994, " Response to Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors," for Nine Mile Point Unit 1.

63 Letter from the NRC to B. R. Sylvia, Executive Vice President - Nuclear, Niagara Mohawk Power Corporation, dated January 13,1995, SER regarding the " Response to Generic Letter (GL) 94-03,

'Intergranular Stress Corrosson Cracking of Core Shrouds in BWRs,' for Nine Mile Point Nuclear l

Station Unit No.1 (TAC No. M90102)."

64 letter from C. D. Terry, Vice President - Nuclear Engineering, Niagara Mohawk Power Corporaison, to the NRC, dated January 6,1995, supplemental response to " Generic letter 94-03,

'Integranular Stress Corrosson Cracking of Core Shrouds in Boiling Water Reactors (TAC No. I M90102)," Nine Mile Point Unit I core shroud repair design. )

65 Letter from the NRC to B. R. Sylvia, Executive Vice President - Nuclear, dated March 31,1995, "Nine Mile Point Nuclear Station Unit No.1 (NMPI), Evaluation of Core Shroud Stabilizer Design (TAC No. M91273).*

11-5 NUREG-1544  ;

l

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66 Letter from J. F. Opeka, Executive Vice President - Nuclear, Northeast Nuclear Energy Company (NNECO), to the NRC, dated August 24,1994, " Millstone Nuclear Power Station, Unit No.1, Response to Generic Letter 94-03."

67 ~ letter from the NRC to J. F. Opeka, Executive Vice President - Nuclear, Northeast Nuclear Energy Company, dated January 4,1995, SER regarding the " Millstone Nuclear Power Station Unit 1 -

Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs* (TAC No.

M90246)." l 68 letter from J. F. Opeka, Executive Vice President, Northeast Nuclear Energy Company, to the NRC, dated July 14,1995, " Millstone Nuclear Power Station, Unit 1, Supplemental Response to Generic ,

imiter M.03."

l 69 - letter from the NRC to J. F. Opeka, Executive Vice President - Nuclear, Northeast Nuciem,- Energy

' Company, dated August 11,1995, " Generic Letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors, Millstone Nuclear Power Station, Unit 1, Supplemental Response (TAC No. M92925)."

I 70 letter from L. H. Waldinger, General Manager, Monticello Nuclear Site, to the NRC, )

dated August 23,1994, "Monticello Nuclear Generating Plant, Docket No. 50-263, License No. DPR-22. Initial (30-Day) Response to NRC Generic Letter 94-03, 'latergranular Stress Corrosion Cracking of Core Shroude in Boiling Water Reactors."

71 letter from the NRC to R. O. Anderson, Director Licensing and Management issues, Northern States Power Company, dated January 20,1995, SER regarding " Generic letter (GL) 94-03, 'Integranular l Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors,' Northern States Power Company (NSP), Monticello Nuclear Generation Plant (MNGP) (TAC No. M90101)."

l 72 letter from W. J. Hill, Plant Manager, Monticello Nuclear Generating Plant, to the N1.C, dated I October 25,1994, "Monticello Nuclear Generating Plant, Docket No. 50-263, License No. DPR-22, i

' Supplemental Response to NRC Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of  !

Core Shrouds in Boiling Water Reactors."

r 73 letter from M. C. Kray, Acting Director - Licensing, Philadelphia Electric Company, to the NRC, dated August 24,1994, " Peach Bottom Atomic Power Station, Units 1 and 2, IJmerick Generating ,

Station, Units 1 and 2, Response to Generic Letter 94-03, *lratergranular Stress Corrosion Cracking ot i Core Shrouds in Boiling Water Reactors."

I 74 letter from the NRC to G. A. Hunger, Jr., Director - Licensing, PEco Energy Company, dated (

February 6,1995, Safety Evaluation Report regarding " Generic Letter (GL) 94-03Platergranular l Stress Corrosion Cracking of Core Shrouds in BWRs,' Peach Bottom Atomic Power Station, l Unit Nos. 2 and 3 (TAC Nos. M90105 and M90106)."  ;

75 letter from G. A. Hunger, Jr., Director - Licensing, Philadelphia Electric Company, to the NRC,

~ dated November 7,1994, " Peach Bottom Atomic Power Station, Unit 2, Supplemental Response to Generic letter 94-03, Summary of Core Shroud laspection Results."

76 GENE-523-176-1293, " Evaluation and Screening Criteria for the Peach Bottom Unit-2 Shroud," dated J

December 13, 1993.

]

NUREG-1544 11-6

)

1 77 letter from G. A. Hunger, Jr., Director - Licerug, PECo Energy Company, to the NRC, dated March 14,1994, " Peach Bottom Atomic Power $tstion, Unit 3, Evaluation of Core Shroud Indications."

78 GENE-523-141-1093, Rev.1, " Evaluation and Screening Criteria for the Peach Bottom Unit-3 Shroud Indications," dated December 13, 1993."

, 79 Ietter M. C. Kray, Acting for G. A. Hunger, Jr., Director - Licensing, Philadelphia Electric Company, to the NRC, dated June 16,1995, " Peach Bottom Atomic Power Station, Unit 3, Submittal

of Inspection Plan in Response to Generic Letter 94-03, 'Intergranular Stress Corrosion Cracking of l'

, Core Shrouds in Boiling Water Reactors'."

80 Internal NRC Memorandum from R. A. Hermann to J. F. Stolz, dated September 25,1995, " Staff Acknowledgement in Regard to Philadelphia Electric Cop . ' s Supplemental Response to Generic Letter 94-03.*

i 81 letter from G. A. Hunger, Director - Licensing, Philadelphia Electric Cornpany, to the NRC, dated j September 16,1994, " Peach Bottom Atomic Power Station, Unit 2, Submittal of Proposed Alternative Repair Plan in Accordance with 10 CFR 50.55a(s)(3)."

82 1.etter from W. A. Josiger, Acting Executive Vice President, New York Power Authority, to the NRC, dated August 24,1994, " James A. FitzPatrick Nuclear Power Plant, Docket No. 50-333 Response to ,

Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs'.* I i

letter from W. J. Cahill, Jr., Executive Vice President - Nuclear Generation, New York Power 83 Authority, to the NRC, dated October 18,1994, " Supplemental Response to Generic letter 94-03,

)

j

'latergranular Stress Corrosion Cracking of Core Shrouds in BWRs'."

84 Letter from William J. Cahill, Jr., Executive Vice President - Nuclear Generation, New York Power Authority, to rne NRC, dated November 30,1994, " Revision to Core Shroud Safety )

Assessment Report."

85 Letter from the NRC to W. J. Cahdl, Jr., Executive Vice President - Nuclear Genemtion, New York Power Authority, dated February 5,1995, Safety Evaluation regarding the " Response to Generic Letter 94-03 for the James A. FitzPatrick Nuclear Power Plant (TAC No. M90092)."

86 letter from W. J. Cahill, Jr., Executive Vice President - Nuclear Generation, New York Power Authority, to the NRC, dated October 21,1994, " Request for NRC Approval of the Fitzpatrick Core Shroud Repair."

87 Ietter from the NRC to W. J. Cahill, Jr., Executive Vice President - Nuclear Generation, New York Power Authority, dated January 5,1995, " Safety Evaluation Regarding the Core Shroud Repair for the James A. FitzPatrick Nuclear Power Plant (TAC No. M90964)."

88 letter from R. D. Machon, Site Vice President, Browns Ferry Nuclear Plant, to the NRC, dated August 23,1994, " Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - Response to Generic letter (GL) 94 Intergranular Stress Corrosion Cracking (IGSCC) of Shrouds In Boiling Water Reactors." ,

89 letter from the NRC to O. D. Kingsley, Jr., President and Chief Nuclear Officer, Tan- Valley Authority, dated January 13,1995, " Browns Ferry Nuclear Plant Units 1,2, and 3, j Safety Evaluation of Response to Generic Ietter 94-03, (TAC Nos. M90081, M90082, and M90083)."

11 7 NUREG-1544

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l 90 I anar from P. S. Salas, Site Licensing Manager, Browns Ferry Nuclear Plant, to the NRC, dated November 18,1994, " Browns Ferry Nuclear Plant (BFN)- Unit 2 - Results of Core Shroud lampactson (TAC No. M90082)."

91 letter from J. P. Pelletier Vice Preandent - Engineenng, Vermont Yankee Nuclear Power i Corporation, to the NRC, dated August 17,1994, " Response to USNRC Generic letter 94-03, ,

'latergranular Stress Corrosion Cracking of Core Shrouds in BWRs'."  ;

92 iamar from the NRC to D. A. Reid, Vice President of Operations, Vermont Yankee Nuclear Power Corporation, dated January 5,1995, " Safety Evaluation for Vermont Yankee Nuclear Power Station Regardag Genenc letter 9443, 'Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors' (TAC No. M90ll4)."

93 Letter from R. G. Byram, Senior Vice President - Nuclear Pennsylvania Power & Light Company, to the NRC, date August 24,1994, "Susquehanna Steam Electric Station, initial Response to Generic f anar 94 03, Docket Nos. 50-387 and 50-388."

94 letter from C. R. Hutchinson, Vice President - Operations, Grand Gulf Nuclear Station, to the NRC, 7 dated August 19,1994, " Grand Gulf Nuclear Station Unit 1. Docket No. 50-416, License No. NPF-29, .

?- ;---- to Genenc letter 94-03." l 95 letter from the NRC to D. L. Farrar, Manager - Nuclear Regulatory Services, Commonwealth Llamaa Company, dated February 16,1995, Safety Evaluation regarding "leselle County Station, Unita 1 and 2 - Response to Generic Letter 94-03, (TAC Nos. M90097 and M90098)."

96 letter from the NRC to G. A. Hunger, Jr., Director - Licensing, PEco Energy Company, dated March 7,1995, Safety Evaluation regarding "Genene Letter (GL) 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs,' Philadelphia Electric Company (PECo), Limerick Unit i I (TAC No. M90099).*

97 Ietter from the NRC to R. G. Bysam, Senior Vice President Nuclear, Pennsylvania Power and Light Company, dated March 23,1995, Safety Evaluation regarding " Generic letter (GL) 94-03, '

'Intergranular Stress Conosion Cracking of Core Shrouds in BWRs,' Pennsylvania Power and Light Company, Susquah==== Steam Electric Station, Units I and 2 (TAC Nos. M90ll2 and M90ll3)." l 98 letter from the NRC to C. R. Hutchinson, Vice President of Operations - Grand Gulf Nuclear  !

Station, Entergy Operations, Inc., d.ated March 29,1995, Safety Evaluation regarding " Generic letter (GL) 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs,' Entergy Operations, l Inc., Grand Gulf Nuclear Station, Unit I (TAC No. M90093).* I 99 Letter from R. G. Byram, Senior Vice Presiden; - Nuclear, Pennsylvania Power and Light Company, to the NRC, dated December 19,1994, "Susquebana Steam Electric Station laterim Response to l Genenc Letter 94-03."

100 letter from the NRC to R. G. Byram, Senior Vice President - Nuclear, Pennsylvania Power and Light Company, dated April 10,1995, "Susquehanna Steam Electric Station, Units I and 2, Interim Response to Generic letter 9443, 'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs (TAC No, M90112 and M90113)."

101 Letter from R. G. Byram, Senior Vice President - Nuclear, Pennsylvania Power and Light Company, to the NRC, dated April 4,1995, "Susquehanna Steam Electric Station, Generic letter 94-03 Interim laspection Report." (Docket 50 387)

NUREG-1544 11-8 se

102 letter from R. G. Byram, Senior Vice President - Nuclear, Pennsylvania Power and Light Co., to the NRC, dated May 25,1995, "Susquakaana Steam Electric Station, Generic Letter 94-03 Final Inspection Results for Unit i Core Shroud." (Docket 50-387) 103 1 meter from the NRC to R. G. Byram, Semor Vice Presulent - Nuclear. Pennsylvania Power and 1.ight Company, dated May 3,1995, '"=; * ---' Steam Electric Station, Unit 1, Staff Acknowledgement in regard to Pennsylvania Power and Light Company's (PP&L) Supplemental Response to Generic Letter (GL) 94-03 (TAC No. M92098)."

104 letter from S.12 Bruns, Vice President - Nuclear Engmeerms, Public Service Electric and Gas Company, to the NRC, dated August 24,1994, " Response to Generic Letter 94-03, 'Intergranular Stress Corremon Cracking of Core Shrouds in Boiling Wa:er Reactors, Hope Creek Generating Station, Docket No. 50-354."

105 Letter from C. D. Terry, Vice President - Nuclear Engineering, Niagara Mohawk Power Corporation, to the NRC, dated August 23,1994, regarding the Nine Mile Unit 2 ' Response to Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds at Boiling Water Reactors.'

106 Letter from J. J. Fisicato, Director - Nuclear Safety, Entergy Operations, lacorporated, to the NRC, dated August 24,1995, " Response to Generic letter 94-03, River Bend Station - 1 Unit 1/ Docket No. 50-458."

]

107 Letter from D. R. Gibson, Senior Vice President - Nuclear Generation, Detroit Edison Company, to the NRC, dated August 24,1994, " Detroit Edison Response to NRC Generic Latter 94-03."

i 108 letter frona J. G. Cook, Vice Presulent, Illinois Power Company, to the NRC, dated August 24,1994,

' Illinois Power's (IP's), Clinton Power Station's (CPS's) Response to Generic letter (GL) 94-03,

- "Intergranular Stress Corromon Cracking of Core Shrouds in Boiling Water Reactors." i 109 letter from R. A. Stratman, Vice Premdent - Nuclear Perry Station, Centerior Service Company, to the NRC, dated August 24,1994, " Perry Nuclear Power Plant, Docket No. 50-440, Response to Generic letter 9443."

110 Imater from J. W. Baker, Acting Assiaenaca Managing Director of Operations, Washington Public Power Supply System, to the NRC, dated August 24,1994, 'WNP 2, Operating License NPF-21 Response to Generic Letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds'.'

111 letter from the NRC to G. A. Hunger, Jr., Director - Licensing, Philadelphia Electric Company, dated March 13,1995, Safety Evaluation regarding ' Generic letter (GL) 94-03, 'Intergranular Stress Conomon Cracking of Core Shiouds in BWRs,' Limerick Generation Station, Unit 2 (TAC No.

M90100)."

112 Internal NRC Memorandum, dated Decennber 13,1994, " Safety Evaluation Reports for Licensee

• 7--- to Genenc letter 94-03." Note that the actual Safety Evaluation Report regarding the is anne en GL 94 43 for Hope Creek Station was issued to Public Service Electric and Gas Cesapany aher September 30,1995, which is beyond the time acope of this NUREG.

b 11-9 NUREG-1544

113 Letter from the NRC to B. R. Sylvia, Executive Vice President - Nuclear, Niagara Mohawk Corporation, dated February 2,1995, Safety Evaluation regarding " Generic letter (GL) 94 03, j

'Intergranular Stress Corrosson Cracking of Core Shrouds in BWRs,' Nine Mile Point Station, Unit 2 l (NMP-2) (TAC No. M90103)." l 114 letter from the NRC to .i. R. McGaha, Jr., Vice President, Entergy Operations, Incorporated, dated February 3,1995, tlafety Evaluation regarding " Generic Letter (GL) 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs,' Entergy Operations, Inc., River Bend Station (TAC {

No. M90111)."

115 latter from the NRC to D. R. Gipson, Senior Vice President - Nuclear Generation, Detroit Edison Company, dated January 24,1995, Safety Evaluation regarding i

• Generic Letter (GL) 94-03, 'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs,' l Detroit Edison Company, Fermi 2 (TAC No, M90091)."

116 Letter from the NRC to R. F. Phares, Director - Licensing, Clinton Power Station, dated February 10,1995, Safety Evaluation regarding " Response to Generic Letter (GL) 94-03,

'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs,' - Clinton Power Station (TAC No. M90086)."

117 Letter from the NRC to D. C. Shelton, Acting Vice President Nuclear - Perry, Centerict Service Company, dated February 10, 1995, Safety Evaluation regarding " Generic letter (GL) 94 03,

'Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs,' Perry Nuclear Plant, Unit No.1 (TAC No. M90107)."

118 latter from the NRC to J. V. Parrish, Vice President - Nuclear Operations, Washington Public Power Supply System, date May 8,1995, " Closeout of Generic letter 94-03, 'Intergranular Stress Corrosion Cracking of Core Shr: reds in BWRs,' for the Washington Public Power Supply System Nuclear Project No. 2 (TAC No. M90115)."

119 General Electric Conviny RICSIL 071, dated November 22,1994, " Top Guide and Core Plate Cracking."

120 Letter from C. Terry, Executive Chairman, BWRVIP Assessment Committee, to the NRC, dated December 23,1994 " Request for Information Regarding the impact of BWR Core Plate and Top Guide Ring Cracking."

121 Imtter from C. Terry, Executive Chairman, BWRVIP Assessment Committee, to the NRC, dated January 3,1995, " Request for Information Regarding the Impact of BWR Core Plate and Top Guide Ring Cracking."

122 General Electric Company SIL 588, dated February 17,1995, " Top Guide and Core Plate Cracking."

123 laternal NRC Memorandum, from B. W. Sheron, Director Division of Engineering, NRR, to A. C.

Thadani, Associate Director for Technology, dated April 25,1995, " Safety Assessment of BWR Core Plate Ring and Top Guide Ring Cracking."

124 NRC Inspection and Enforcement Bulletin (IEB) 80-07, dated April 4,1980, "BWR Jet Pump l Assembly Failure."  ;

l 125 NRC Information Notice 93-101, dated December 17,1993, " Jet Pump Hold-down Beam Failure." i I

NUREG-1544 11-10

l 1

126 GE SIL No. 330, dated June 1980, " Jet Pump Beam Cracks." l 127 NRC NUREG/CR 3052, dated November 30,1984, " Closeout of IE Bulletin 80-07: BWR Jet Pump Assembly Failure.'

128 GE SIL No. 462, dated February 1,1988, " Shroud Support Access Hole Cracks."

129 GB SIL No. 462, Supplement 1, dated February 22,1989, " Shroud Support Access Hole Cracks."

130 GE SIL No. 462, Supplement 2, dated August 1990, " Shroud Support Access Hole Cracks."

131 GE SIL No. 462, Supplement 2, Revision 1, dated December 31,1990, " Shroud Support Access Hole Cracks." ,

i 132 GE SIL No. 462, Supplement 3, dated June 8,1992, " Radial Cracking in Creviced Inconel 600 AHC Weidments."

133 NRC Information Notice 8843, dated February 2,1988, ' Cracks in Shroud Support Access Hole Covers."

134 NRC Information Notice 92-57, dated August i1,1992, " Radial Cracking of Shroud Suppo 1 Access Hole Cover Welds."

11 11 NUREG-1544 l

APPENDIX A LIST OF BWROG MEMBERS AND BWRVIP SUBCOMMITTEES BWROG Members BWRVIP Subcommittees

• Boston Edison Company BWRVIP Chairman Thomaa Beckham  !

Carolina Power & Light Company Southern Nuclear Operating Corp. l l

Centerior Energy, incorporated BWRVIP Vice Chairman 1 Carl Terry Commonwealth Edison Company Niagara Mohawk Power Company

)

1 Detroit Edison Company lategration Subcomnuttee Entergy Operation, incorporanal Executive Chairman Twhnical Chairman John Hosmer Vaughn Wagner Georgia Power Company Commonwealth Edison CP&L General Public Utahties inspection Subcommittee Illinois Power Company Executive Chairman Technical Chairman IES Utilities,Inc. Robert Keaten Steve Leonard General Public Utilities Niagara Mohawk Nehmska Public Power Diurw1 Niagara Mohawk Power Corpwation Assessment Subcommittee Northeast Nuclear Energy Cawnpany Executive Chairman Technical Chairman Carl Terry Robin Dyle Northern States Power Company Niagara Mohawk Power Co. Southern Nuclear Pennsylvania Power & bght Company Mitigation Subcommittee Philadelphia Electric Company Executive Chairman Technical Chairman Power Authority of the State of New York Vacant John Wilson PSE&G Public Services Electric and Gas Company Repair Subcommittec Tennessee Valley Authority Executive Chairman Technical Chairman Vermont Yankee Nuclear Power Corporation William Campbell Bruce McLeod Carolina Power & Light Georgia Power Co.

Washington Public Power Supply System

• As of January 3,1996 A-1 NUREG-1544

l l

l l

APPENDIX B i

PLANT-SPECIFIC BWR l CORE SHROUD SUMMARIES l

l B-1 NUREG-1544 t

i I

I

PILGRIM NUCLEAR POWER PLANT DATA SHEET l

LICENSEE: Boston Edison Comoany (BEco) j PLANT NAME: Pilerim Nuclear Power Station (PNPS)

DOCKET NO.: 50-293 LICENSE NO.: DPR-35 BWRVIP CATEGORY GROUPING: Cateeorv "C" SIIROUD FABRICATION DATA:

SilROUD MANUFACTURER: P.F. Avery Com.

SilROUD SilELL CONSTRUCTION MATERIALS: ASTM A240-63 Tvoe 304 stainless steel olate CARBON CONTENT RANGE SIIELL SECTIONS: 0.045 % C - 0.54% C SIIROUD RING CONSTRUCTION MATERIALS: ASTM A240-63 Tvoe 304 stainless steel clate CARBON CONTENT R ANGE SHROUD RINGS: 0.054% C - 0.056% C WELDING DATA:

INITIAL PASS TECIINIQUE: Flance welds by SM AW: other shroud ved. & cire. welds hv SAW INITIAL PASS WELD MATERIAL: E-308 electrodes for SM AW: ER-308L filler for SAW SUBSEQUENT PASS TECilNIQUE: Flance welds by SM AW: other shroud vert. & cire. welds by SAW SUBSEQUENT PASS WELD MATERIAL: E-308 electrodes for SM AW: ER-308L filler for SAW OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.53 uS/cm ELAPSED TIME AT POWER OPERATION: > 12.6 EFPY by F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aurust 27.1994 BASIS FOR ACCEPTING JCO: Generic Safety Assessment DATE OF CORE SHROUD INSPECTION / REPAIR: RFO No.10. April 1995 NRC SERs: (1) SER recanime BEco's response to GL 94-03. November 28.1994 (2) SER recardmc BECo modification of the PNPP core shroud. May 12.1995 REMARKS: BECo optsi to implement a modification of the PNPP core shroud in lieu of performine comprehensive core shroud examinations. Modification uses the GE tie rod assembly desien. This desien provides a redundant load oath for the shroud under normal operatine. transient and costulated desien basis accident conditions riven the occurrence of a 360* throuchwall failure of a circumferential weld. Modification of the PNPP core ahroud is desiened to maintain the structural intecrity of the PNPP core shroud durine subseauent operatine cycles. Staff soproved the PNPP core shroud modification desien in May 1995.

NUREG-1544 B-2 l

1

BRUNSWICK UNIT 1 DATA SIIEET i

LICENSEE: Carolina Power and Licht Company (CP&L) l PLANT NAME: Brunswick Steam Electric Plant. Unit I (BR-1)  !

DOCKET NO,: 50-325 LICENSE NO.: DPR-71 BWRVIP CATEGORY GROUPING: Cateeory "C" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Shin Building and Dry Dock Co.

SilROUD SHELL CONSTRUCTION MATERIAL: A/SA 240 Tyne 304 stainless steel rolled olates CARBON CONTENT RANGE SIIELL SECTIONS:_ 0.048% C - 0.064% C SHROUD RING CONSTRUCTION MATERIAL: A/SA 240 Tvoe 304 stainless steel rolled plates CARBON CONTENT RANGE SilROUD RINGS: 0.063% C - 0.078% C WELDING DATA:

INITIAL PASS TECHNIQUE: SM AW or S AW methods INITIAL PASS WELD MATERIAL: E308 SUBSEQUENT PASS TECHNIQUE: SMAW or SAW methods SUBSEQUENT PASS WELD MATERIAL: E308 OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.578 uS/cm ELAPSED TIME AT POWER OPERATION: > 9.8 EFPY hv F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 24.1994 BASIS FOR ACCEPTING JCO: Results of F1993 flaw evaluation and implementation of shroud repair DATE OF CORE SHROUD INSPECTION / REPAIR: Inanect. S1993. Renair W1993/1994. Reinsoection SP1995 NRC SERs: (1) SER regardine BR-1 core shroud repair January 14. 1994 (2) SER recardine CP&L's response to GL 94-03. January 3.1995 REMARKS: Initial VT-1 and UT ex==ination of BR-1 core shroud in S1993. in accordance with reces..;=Aa' ions in GE RICSIL 054. Results of examinations indicated larne scale (360") erackine at the H3 weld. Results also indicated a significant amount of crackine at welds 111 and H2. and minor crackine at weld H4. CP&L ingdemented a nxxlification of the BR-1 shroud usine a series of mechanical clamos around the H2 and H3 welds.

The clamas are desirned to orovide a redundant load oath around the H2 and H3 welds. Modification A=ien approved by staffin January 1994. CP&L oerformed a second inspection (UT) of the BR-1 shroud in April 1995.

No new crack initiation discovered: erowth of existine indications minimal.

B-3 NUREG-1544

BRUNSWICK UNIT 2 DATA SIIEET LICENSEE: Caralina Power and Linht Coma =av (CP&L)

PLA?E NAME: Bnaswick Steam Electric Plant. Unit 2 (BR-2)  :

DOCKET NO.: 50-324 i LICENSE NO.: DPR-62 BWRVIP CATEGORY GROUPING: Catenorv 'C' SHROUD FABRICATION DATAt I SHROUD MANUFACTURER: Sun Shin Building and Dry Dock Co.

SHROUD SHELL CONSTRUCTION MATERIAL: A/SA 240Tvoe 304 rolled plates CARBON CONTENT RANGE SHELL SECTIONS: 0.046% C - 0.061 % C SHROUD RINO CONSTRUCTION MATERIAL: A/SA 240 Tyne 304 rolled plates ,

I CARBON CONTENT RANGE SHROUD RINGS: 0.047 % C - 0.067 % C WELDING DATA:

INITIAL PASS TECHNIQUE: SMAW or SAW methods INITIAL PASS WELD MATERIAL: E308L electrrxle or ER308L filler as nnpropriate SUBSEQUENT PASS TECHNIQUE: SMAW or SAW methods SUBSEQUENT PASS WELD MATERIAL: E30RL electrode or ER308L filler as nopropriate OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.714 uS/cm ELAPSED TIME AT POWER OPERATION: > 10.1 EFPY hv F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aunust 24.1994 i BASIS FOR ACCEPTINO JCO: Results of BR-1 S1993 shroud exams bounding for BR-2. also reneric assessment DATE OF CORE SHROUD INSPECTION / REPAIR: S1994: Next inspection scheduled for S1996 l NRC SERs: SER renardine response to GL 94-03. January 3.1995 l REMARKS: CP&L implemented a repair of the BR-2 core shroud durine the SP1994 RFO. Rengir desien  ;

encomnassed circumferential welds H2 and H3. Reoair desian was the same one as was used to modify the BR-1 i core shroud in the Wl994/1995. CP&L also performed examinations of BR-2 shroud welds that were not coverg4 by the desien modification. These examinations included enhanced VT-1 inspections of the Hl . H2. H4. H5. H6a.

H6b. and H7 circumferential welds. and a UT inspection of the H4 wnld. Only minor crackins was detected as a result of the examinations. Maximum circumferential crack length was 0.35 m (13.6 in) at weld H4 (an 0.28-m ill-ini erack was detected at weld HS: all other indications < 0.05 m i2 inl lone). Flaw evaluatign iustifies current operation.

NUREG-1544 B-4

i 1

l l

1 DRESDEN UNIT 3 DATA SHEET  !

LICENSEE: Commonwealth Edison Comnany (Comed)

PLANT NAME: Dresden Nuclear Power Plant. Unit 3 (DR-3) l DOCKET NO.: 50-249 LICENSE NO.: DPR-25 BWRVIP CATEGORY GROUPING: Catenory "C" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Binohnm Willi = matte SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tyne 304 SS rolled clate CARBON CONTENT RANGE SHELL SECTIONS: 0.044% C - 0.063 % C (rance for whole shroud)

SHROUD RING CONSTRUCTION MATERIAL: A 240 Tvoe 304 SS rolled plate CARBON CONTENT RANGE SHROUD RINGS: 0.044% C - 0.063 % C (rance for wholn shroud) l WELDING DATA:

INITIAL PASS TECHNIQUE: Techniaue not specified in the reosoonse to GL 94-03 INITIAL PASS WELD MATERIAL: ASTM Tvoe E308 electrode or ER308 filler material

• SUBSEQUENT PASS TECilNIQUE: Techniaue not specified in the resconse to GL 94-03 '

SUBSEQUENT PASS WELD MATERIAL: ASTM Tvoe E308 electrode or ER308 filler material OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER

- THE FIRST FIVE YEARS OF OPERATION: 0.399 uS/cm ELAPSED TIME AT POWER OPERATION: > - 15 EFPY by F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aurust 23.1994 '

BASIS FOR ACCEPTING JCO: Flaw evaluation of DR-3/OC-1 cort shrouds DATE OF CORE SHROUD INSPECTION / REPAIR: Renair schedule <l for F1996 NRC SERs: (1) SER reeardine DR-3/OC-1 core shrouds. July 21.1994 (2) SER renardine resconse to GL 94-03. January 31.1995 (3) SER renardine confirmation of DR-3/OC-1 flew evaluation August 16.1995 i

REMARKS: ComFA cerformed a u,su,r '- = rive inspection of the DR-3 core shroud in SP1994. Insoections inci ~ tad -kaarad VT-1 e===ia=tions of welds Hi - H7 and UT examinations of welds H2. H5. H6. and H7.

Pa==1tm of ComF11's examinations indicated the cresence of extensive circumferential crackinc in the H5 weld.

Cracks tra=*ad as ex*=Aiao 360' around the weld. The NRC performed an independent flaw evaluation of the H5 weld. NRC daiarmiaad that the H5 weld should have sufficient re==inine lir==a=*s in the weld to iustify an additional 15 maank of hot operation of the DR-3 unit. The NRC also determined that any cracking detected at  !

other circumferential welds would be bonad~i by that at the H5 weld.

l l

4 l

l B-5 NUREG-1544 I

! I

QUAD CITIES UNIT 1 DATA SHEET LICENSEE: Commonwealth Edison Comnany (Comed)

PLANT NAME: Ouad Cities Nuclear Power Plant. Unit 1 (OC-1)

DOCKET NO.: 50-254 LICENSE NO.: DPR-29 BWRVIP CATEGORY GROUPING: Catecorv "C" SliROUD FABRICATION DATA:

SHROUD MANUFACTURER: Bincham Williamette SIIROUD SHELL CONSTRUCTION MATERIAL: A 240 Tyne 304 SS rolled nlate CARBON CONTENT RANGE SHELL SECTIONS:_0.044% C - 0.063% C (rance for whole shroud)

SHROUD RING CONSTRUCTION MATERIAL: A 240 Tyne 304 SS rolled plate CARBON CONTENT RANGE SHROUD RINGS: 0.044% C - 0.063 % C (rance for whole shroud)

WELDING DATA:

INITIAL PASS TECHNIQUE: Techniaue not specified in the resn<mse to GL 94-03 INITIAL PASS WELD MATERIAL: ASTM Type E308 electrode or ER308 filler material SUBSEQUENT PASS TECHNIQUE: Techniaue not specified in the resnonse to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: ASTM Tyne E308 electrode or ER308 filler material OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.377 uS/cm ELAPSED TIME AT POWER OPERATION: > 16.0 EFPY by F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aucust 23.1994 BASIS FOR ACCEPTING JCO: Flaw evaluation of DR-3/OC-1 core shrouds DATE OF CORE SHROUD INSPECrlON/ REPAIR: Insnection performe<1 SP1994 NRC SERs: (1) SER recardine DR-3/OC-1 core shrouds. July 21.1994 (2) SER recardine response to GL 94-03. January 31.1995 (3) SER recardinc confirmation of DR-3/OC-1 flaw evaluation. Aucust 16.1995 REMARKS: Comed nerformed a comprehensive insnection of the OC-1 core shroud in SP1994. Inspections included enhanced VT-1 examinations of welds HI - H7 and UT examinations of welds H2. H5. H6. and H7.

Realts of Comed's examinations indicated the presnee of extensive circumferential crackinc in the H5 weld.

Cracks treated as extendinc 360* around the weld. The NRC performed an independent flaw evaluation of the H5 weld. NRC determined that the H5 weld should have sufficient remaininc licaments in the weld to iustify g .

gl.ditional 15 months of hot operation of the OC-1 unit. 'the NRC also determined that any crackine detected at other circumferential welds would be bounded by that at the H5 weld.

NUREO-1544 B4

)

1 l

l DRESDEN UNIT 2 DATA SIIEET i

LICENSEE: Commonwealth Edison Comn=nv (Comed)

PLANT NAME: Dresden Nuclear Power Plant. Unit 2 (DR-2)

)

DOCKET NO.: 50-237  !

LICENSE NO.: DPR-19 BWRVIP CATEGORY GROUPING: Categorv "C' i 1

SHROUD FABRICATION DATA.

SHROUD MANUFACTURER: Bineham Williamette SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tvoc 304 SS rolled plate i

CARBON CONTENT RANGE SHELL SECTIONS: 0.044% C - 0.063 % i (rance for whole shroud) l SIIROUD RING CONSTRUCTION MATERIAL: A 240 Type 304 SS rolled niate CARBON CONTENT RANGE SHROUD RINGS: 0.044% C - 0.063% C (rance for whole shroud)

WELDING DATA:

INITIAL PASS TECHNIQUE: Techniaue not specified in response to GL 94-03 INITIAL PASS WELD MATERIAL: ASTM Tvoe E308 electrode or ER308 filler material SUBSEQUENT PASS TECHNIQUE: Techniaue not specified in resnonse to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: ASTM Tvoe E308 electrode or ER308 filler material OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER Tile FIRST FIVE YEARS OF OPERATION: 0.299 uS/cm ELAPSED TIME AT POWER OPERATION: > 17.1 EFPY by F1994 GENERIC LETTER INFORMATION:  !

DATE GL 94-03 RESPONSE: August 23.1994 l BASIS FOR ACCEPTING JCO: Results of DR-3/OC-1 flaw evaluations boundine for DR-2/OC-2 l DATE OF CORE SHROUD INSPECTION / REPAIR: Renair scheduled for F/W1995

~

NRC SERs: (1) SER rerardine the Comed response to GL 94433 for DR-2/OC-2. January 31.1995 REMARKS: Comed opted to imnlement a modification of the DR-2 core shroud in lieu of cerfornung comnrehensive core shroud ex==in=tions. The desien involves installation of GE desiened tie rod assemblies around the OD of the shroud. Tie rod assemblies are designed to orovide an alternate load hearine canability for the shroud durine normal operatine. transient and costulated desien basis accident conditions. riven the occurrence of a 360" throurh-wall failure of a circumferential weld. Core shroud modification desien currently under review by the NRC staff.

B7 NUREG-1544

QUAD CITIES UNIT 2 DATA SHEET LICENSEE: Com-wealth Edison C-v (Comed)

PLANT NAME: Ouad Cities Nuclear Power Plant. Unit 2 (OC-2)

DOCKET NO.: 50-265 l LICENSE NO.: DPR-30 BWRVIP CATEGORY GROUPING: Cateeorv "C" SHROUD FABRICATION DATA: i SHROUD MANUFACTURER: Bineham Willi =-ne J SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tvoe 304 SS rolled clate '

. CARBON CONTENT RANGE SHELL SECTIONS: 0.044 % C - 0.063% C (ranee for whole shroud)

SHROUD RING CONSTRUCTION MATERIAL: A 240 Type 304 SS rolled plate CARBON CONTENT RANGE SHROUD RINGS: 0.044% C - 0.063 % C france for whole shroud)

WELDING DATA:

INITIAL PASS TECHNIQUE: Techniaue not anecified in response to GL 94-03 INITIAL PASS WELD MATERIAL: ASTM Tvoe E308 electrode or ER308 filler material SUBSEQUENT PASS TECHNIQUE: Techniaue not specified in response to GL 9443 SUBSEQUENT PASS WELD MATERIAL: ASTM Tvoe E308 electrode or ER308 filler material  !

OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.377 uS/cm ELAPSED TIME AT POWER OPERATION: > 16.1 EFPY by F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 23.1995 BASIS FOR ACCEPTING JCO: P-I s tof DR-3/OC-1 flaw evaluation boundine for DR-2/OC-2 DATE OF CORE SHROUD INSPECTION / REPAIR: Repair implemented SP/S1995 NRC SERs: (1) SER renardine the response to GL-94-03 for DR-2/OC-2. January 31.1995 (2) OC-2 core shroud repair (modification) SER. June 8.1995 REMARKS: Comed ooted to implement a modification of the OC-2 core shroud in lieu of performine comprehensive core shroud examinations. The desien involves installation of GE desiened tie rod assemblies around the OD of the shroud. Tie sud assemblies are desiened to orovide an alternate load bearine capability for the shroud durine normal operatine. transient and costulated desien basis accident conditions. civen the occurrence of a 360*

throuch-wall failure of a circumferaatial weld. The NRC anoroved the modification desien for the OC-2 core shroud on June 8.1995.

NUREG-1544 B-8

l l

I OYSTER CREEK DATA SHEET l l

l l

LICENSEE: General Public Utilities (GPU)

PLANT NAME: Oyster Creek Nuclear Generation Station (OCNGS)

DOCKET NO.: 50-219 LICENSE NO.: DPR-16 BWRVIP CATEGORY GROUPING: Cateeorv 'C' i l

SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: P.F. Avery SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Type 304 SS rolled niates CARBON CONTENT RANGE SI1 ELL SECTIONS: 0.042 % C - 0.062 % e SHROUD RING CONSTRUCTION MATERIAL: A 240 Tyne 304 SS rolled plates CARBON CONTENT RANGE SHROUD RINGS: 0.056% C - 0.064% C WELDING DATA:

INITIAL PASS TECHNIQUE: Information not provided in response to GL 94-03 INITIAL PASS WELD MATERIAL: Information not provided in resnonse to GL 94-03 SUBSEQUENT PASS TECHNIQUE: Information not provided in response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Information not nrovided in resnonse to GL 94-03 OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.526 uS/cm ELAPSED TIME AT POWER OPERATION: > 15.5 EFPY by F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aueust 24.1994 BASIS FOR ACCEPTING JCO: Generic Safety Assessment and Short time to F1994 RFO DATE OF CORE SHROUD INSPECTION / REPAIR: Inspection. Repair durine F1994 RFO NRC SERs: (1) SER rerardine the OCNGS core shroud repair. November 25.1904 (2) SER reearding the responw to GL 94-03 for OCNGS. February 23.1995 REMARKS: GPU oerformed comprehensive UT (w/ some VT-1) inspections of the OCNGS core shroud durine the F1994 RFO. Inspection coverares ranced from 12%-27% for welds HI-H3. and 31%-49% for welds H4-H6B. Welds H7 and H8 were not included in the insnections due to the presence of additional structural brackets in the shroud desien. Inspections of the H4 revealed the presence of substantial crackine (taken to be 360')

in the welds HAZ. OCNGS opted to perform a corrective renair of the OCNGS core shroud prior to restart of the OCNGS unit. The repair involved the installation of a number of MPR41esiened tie rod assemblies around the outer circumference of the shroud. GPU's core shroud modification desien for the OCNGS was annroved by the staff prior to restart of the unit.

B-9 NUREG-1544

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

HATCH UNIT 1 DATA SHEET LICENSEE: Georeis Power C-av (GPC)

PLANT NAME: Edwin I. Hatch Nuclear Plant. Unit I (HAT-1)

DOCKET NO.: 50-321 LICENSE NO.: DPR-57 BWRVIP CATEGORY GROUPING: Catenorv 'C' SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Shm Buildaar and Dry Dock Co.

SHROUD SHELL CONSTRUCI'lON MATERIAL: A/SA 240 Tyne 304 SS rattari al=*=

CARBON CONTENT RANGE SHELL SECTIONS: tar-*iaa not orovided in the ra aa== to GL 94-03 SHROUD RING CONSTRUCTION MATERIAL: A/SA 240 Tyne 304 SS rolled plate CARBON CONTENT RANGE SHROUD RINGS: Information not provided in the ramaaa_= to GL 94-03 WELDING DATA:

INITIAL PASS TECHNIQUE: Infor-a*:aa not orovided in the .-:+ == to GL 94-03 I INITIAL PASS WELD MATERIAL: Information not orovided in the ra aaa- to GL 94-03 SUBSEQUENT PASS TECHNIOUli: Infor==*iaa not provided in the i-- = to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Information not provided in the e- to GL 94-03 ,

l OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER  ;

THE FIRST FIVE YEARS OF OPERATION: 0.41I uS/cm j ELAPSED TIME AT POWER OPERATION: > 12.8 EFPY by F1994  !

GENERIC IETTER INFORMATION:

DATE GL 94-03 RESPONSE: Au=unt 24.1994 BASIS FOR ACCEPTING JCO: Generic Safety and A-- -- - =" and Short time duration until the F1994 RFO DATE OF CORE SHROUD INSPECTION / REPAIR: Renair. "-r-i-:='+-/ October 1994 NRC SERs: (1) SER renardiaa the HAT-1 core shroud mir. "-x^ -- -Mi 30. 1994 -

(2) SER ra== dian GPC's resnonne to GL 94-03 for HAT-1/H AT-2. February 23.1995 l

t REMARKS: GPC opted to nerform := a_,;ive renair of the HAT-1 com shroud durina the "-M-:--_-Ni 1994  !

RFO in lieu of ca===-l====3ve core shroud ===-la=* ions. H AT-1 shroud modification desien submitted to the staff ,

on * - ' - 1 1994. 'The da=i== involves a-*- Id= of GE da=3ened tie rod --

l

' lies aio-.d the OD of the abroud. Tae rod assembhes are da=i aad to provide an alternate load bearina c=a hility for the shroud durine

_ u, . {

__. and --- ^- ' -"da Len basis cidaae n nadifiana. eiven the --_ - .., s of a 360* throuah- l wall f=it-e of a cir_ ' atial ;. 'd. Desien --,.uved by the NRC prior to restart of the HAT-1 unit.

i i

i NUREG-1544 B-10 T

1 i

+- -. -t?

, HATCH UNIT 2 DATA SHEET d

, LICENSEE: Georris Power Company (GPC) l PLANT NAME: Edwin I. Hatch Nuclear Plant. Unit 2 (HAT-2)

, DOCKET NO.: 50-366 LICENSE NO.: NPF-5 BWRVIP CATEGORY GROUPING: Catenorv *C" SHROUD FABRICATION DATA:

SHROUD M ANUFACTURER:,,lun Shin BuilAiaa and Dry Dock Co.

SHROUD SilELL CONSTRUCTION MATERIAL: A/SA 240 Tyne 304 SS rolled nlate CARBON CONTENT RANGE SHELL SECTIONS: Infom*iaa not orovided in the r=-= to GL 94-03 SHROUD RING CONSTRUCTION MATERIAL: A/SA 240 Tvoe 304 SS rolled plate CARBON CONTENT RANGE SHROUD RINGS: Information not orovided in the resnonse to GL 94-03 WELDING DATA:

INITIAL PASS TECHNIQUE: Information not provided in the ramanne to GL 94-03 INITIAL PASS WELD M ATERIAL: Information not nrovided in the rannan= to GL 94-03 SUBSEQUENT PASS TECIINIQUE: Infom*iaa not orovided in the rm= to GL 94-03 SUBSEQUENT PASS WELD M ATERIAL: Information not provide in the response to GL 94-03 OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS Of OPERATION: 0.459 nS/cm ELAPSED TIME AT POWER OPERATION: > 10.0 EFPY by F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aunust 24.1994 BASIS FOR ACCEPTING JCO: Insnection results and flaw evaluations of the HAT-2 shroud durina SP1994 RFO DATE OF CORE SHROUD INSPECTION / REPAIR: Insnection SP1994: mir =eliadalad for F1995 NRC SERs: (1) SER renardine GPC's response to GL 94-03. February 23.1995 (2) SER regardine the HAT-2 core shroud raa ir. *-a --'er 25.1995 REMARKS: GPC nerformed UT insoections of the H1-H4 cin i.- ^:=1 welds durine the HAT-2 SP1994 RFO:

GPC also nerformed uartial VT-1 e=- l--*ia== of =li H5-H8. UT *= ^^ ions of ==dble oce:aan of the H2 weld 3 dia-sad the or- e of cr=eiri== totaline 5.54 m (218 in) in !=-i. with the le - !=Eh t"--

= 159 i= 8 a- in ta==th. UT ia a-c*sa== of the HAT-2 shroud also rev t d a =taae amount of cr=caria= at ;;li Hl. H3. and H4 welds. Partial VT-1 ====ia-*1aan of the lower shroud welds wem =a==tive for flaw i-dia=*iaam

'llie r- *" of flaw evala=*ia== of the HAT-2 d.iesd iustified oner=sia= of the =aie to the F1995 RFO. P-ir daeann will involve the installation of a number of GE-desian tie rod assembhos around the outer circumference of the shroud na.=i, da.i.= anoroved by the NRC in E-:- -- ' = 1995.

B-11 NUREG d544

1 l

)

DUANE ARNOLD ELECTRIC CENTER DATA SHEET LICENSEE: IES Utilities. lac. (IES)  ;

PLANT NAME: D - Amold Essev C=^-- (DAEC) l DOCKET NO.: 50-331 LICENSE NO.: DPR-49 i

)

BWRVIP CATEGORY GROUPING: Cateeorv 'C' SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Biaak = Will ---**a SHROUD SHELL CONSTRUCTION MATERIAL: Tvoe 304L asaintass steel CARBON CONTENT RANGE SHELL SECTIONS: Infor==sina not provided in the resoonse ta_GL 94-03 SHROUD RING CONSTRUCTION MATERIAL: Tvoe 304L atainless steel CARBON CONTENT RANGE SHROUD RINGS: Information not provided in the response to GL 94-03 l WELDING DATA:

INITIAL PASS TECHNIQUE: Information not provided in the resnora,e so GL 94-03 '

INITIAL PASS WELD MATERIAL: Infomeian not provided in the response to GL 94-03 SUBSEQUENT PASS TECHNIQUE: Information not provided in the response to GL 94 03 SUBSEQUENT PASS WELD MATERIAL: Information not provided the in response to GL 94-03 )

i

- OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER 1 THE FIRST FIVE YEARS OF OPERATION: near mid ranee conductivity for the industry (= 0.3 nS/cm) l' ELAPSED TIME AT POWER OPERATION: > 13.5 EFPY by F1994 GENERIC LETTER INFORMATION:  :

DATE GL 94-03 RESPONSE: Aunust 24.1994 i BASIS FOR ACCEITING JCO: Generic safety assessment l DATE OF CORE SHROUD INSPECTION / REPAIR: Cc. .md.asive UT examinations durine W/SP1995 RFO

]

NRC SERs: SER renardine renoonse to GL 94-03. March 1.1995 l l

l REMARKS: Cc, awed.c=ive UT e== iaations of DAEC core shroud durine W/SP1995 RFO. Examinations j covered accessible nortions of H1-H7 from the OD. IES nerformed the UT examinations using the GE OD l Traekae. with 45' ka- wave. 60' loneitudinalwave and creen UT tuhai=_m. Weld coverane raneed from 52.4%

at weld H2 to 78.8% at weld H4, No flaw indications evident as a result of the examinations nerformed durine the W1995 RFO.

NUREG-1544 B 12

NINE MILE POINT UNIT 1 DATA SIIEET LICENSEE: Nine=rs Mohawk Power Corooration (NMPC)

PLANT NAME: Nine Mile Point Station. Unit 1 (NMP-1)

DOCKET NO.: 50-220

{, LICENSE NO.: DPR-63 l

5 BWRVIP CATEGORY GROUPING: Caterorv 'C"

] SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: P.F. Averv l SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tvoe 304 SS clate 1 CARBON CONTENT RANGE SHELL SECTIONS: 0.042% C - 0.062% C i

SHROUD RING CONSTRUCTION MATERIAL: A 240 Tvoe 304 SS olate (shroud supoort rinc: SA 336 tvoe F8 CARBON CONTENT RANGE SHROUD RINGS: 0.056% C - 0.064% C WELDING DATA:

l INITIAL PASS TECHNIQUE: SAW l INITIAL PASS WELD MATERIAL: A-371 Tvoe ER308 Filler (5% minimum ferrite content) l SUBSEQUENT PASS TECHNIQUE: SAW )

1 SUBSEQUENT PASS WELD MATERIAL: A-371 Tvoe ER308 Filler (5% minimum ferrite content) I i j j OPERATIONAL DATA: l AVERAGE CONDUCTIVITY VALUE OVER i THE FIRST FIVE YEARS OF OPERATION: 0.457 nS/cm ELAPSED TIME AT POWER OPERATION: > 14.4 EFPY hv F1994 i

i- GENERIC LETTER INFORMATION:

l DATE GL 94-03 RESPONSE: Aurust 23.1994 BASIS FOR ACCEPTING JCO: Generic Safety Asms=at l DATE OF CORE SHROUD INSPECrlON/ REPAIR: Renair of NMP-1 SPl995 RFO NRC SERs: (1) SER rerardine the response to GL 94-03 for NMP-1. January 13.1995 (2) SEN renarding the NMP-1 core shroud repair. March 31.1995 i

i 3 REMARKS: NMPC wif6rir.ed a ore-emotive recair (modification) of the NMP-1 core shroud in SP1995 RFO, This modification was is rAr rt.:d in lieu of nerformina in depth core shroud examinations. NMPC submitted the modification desien on Jan. 6.1995. The modification desien involves the use of tie rod assemblies designed by the GE Comnany. The tie rod assemblies are desianed to provide an alternate load oath for the core shroud and are "=3 to --- the worst case laadian conditians durine normal operatine. transient. and whilanad des sc i '- t eWhm riven the occu m.ce of a 360* throuah-wall failure of a circu=,..tial weld. The NRC anosoved the modification for implementation on March 31.1995.

i i

a B-13 NUREG-1544

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

l l

l COOPER NUCLEAR STATION DATA SHEET l l

4 LICENSEE: Nebr==k= Public Power District (NPPD)

. PLAPft' NAME: Caaaac N=el-e **=*1a= (CNS) I DOCKET NO,: 50-298 ,

LICENSE NO.: DPR-46

]

BWRVIP CATEGORY GROUPING: c-a--arv 'C' i

SHROUD FABRICATION DATA:  !

SHROUD M ANUFACTURER: * * -- Willb ^^

SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tyne 304 SS hot 1 '!='. ---- * ' ' -

CARBON CONTENT RANGE SHELL SECTIONS: 0.043% C - 0.068% C -

SHROUD RING CONSTRUCTION MATERIAL: A 240 Tyne 304 SS hot rallad ---- '-4 '

• CARBON CONTENT RANGE SHROUD RINGS: 0.052% C - 0.058% C j

, i WELDING DATA:  :

INITIAL PASS TECHNIQUE: SAW _

l

. INITIAL PASS WELD MATERIAL: ER308 Fillar SUBSEQUENT PASS TECHNIQUE: SAW SUBSEQUENT PASS WELD MATERIAL: ER308 Filler )

i OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER 1

. THE FIRST FIVE YEARS OF OPERATION: 0.188 uS/can  !

ELAPSED TIME AT POWER OPERATION: > 14.6 EFPY by F1994 l GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aurunt 26.1994 BASIS FOR ACCEPTING JCO: e - - - ,c e '_ # A - ^: ec==!=1 with ki=*arv of cl--- r=:~ w-:'- '

DATE OF CORE SHROUD INSPECTION / REPAIR: Comprehensive UT Inspection scheduled for October 1995 NRC SERs: ( 1 ) SE R ,,,,,.a... , _ _ _ _- - - to GL 94-03. April 12.1995 (2) Ach=awledea====e letter of core shroud a aae* ion =c==. "- ^ --*- :20.1995

~ REMARKS: Last lie-- with a r=*a-orv 'C' niant to i==aae* its core shroud This --*:aa will e==al-*= the initial est ofinspections/ repairs Categorv "C" shrouds. The proposed inspection scope for CNS shroud is consistant

. i ths -- - ^ of -*- = c ^--- ,, v 'C' '==---

:aa --- = that have been whoitted to the NRC. i== =::aa u---

annroved by NRR.

NUREG-1544 B-14 Wrr -" -'wwvv =

bT 'S " "

• j

\

, i s

1 l MILLSTONE UNIT 1 DATA SIIEET

4 i LICENSEE: Northeast Nuclear Enerry Comaaav (NNECO)

PLANT NAME: Mill =*- Unit I (MS-1) I DOCKET NO.: 50-245 LICENSE NO.: DPR-21 i

BWRVIP CATEGORY GROUPING: Catenorv 'C'

! I 1 SHROUD FABRICATION DATA:

, SHROUD MANUFACTURER: P.F. Averv

SHROUD SHELL CONSTRUCTION MATERIAL

Tyne 304 SS nlate i j CARBON CONTENT RANGE SHELL SECTIONS: infw iaae not nrovided in the i-w----- to GL,94-03 j SHROUD RING CONSTRUCTION MATERIAL: Tyne 304 SS nlate i

j CARBON CONTENT RANGE SHROUD RINGS: Information not nrovided in the ra=- a to GL 94-03 j WELDING DATA:

INITIAL PASS TECHNIQUE: Infor==*ina not nrovided in the ramana=a to GL 94-03 I INITIAL PASS WELD MATERIAL: Information not nrovided in the r==aaa- to GL 94-03

)

l SUBSEQUENT PASS TECHNIQUE: Information not nrovided in the ra aan- to GL 94-03 '

SUBSEQUENT PASS WELD MATERIAL: Information not nrovided in the ra aaa- to GL 94-03 OPERATIONAL DATA:

1 AVERAGE CONDUCTIVITY VALUE OVER l j THE FIRST FIVE YEARS OF OPERATION: < < 0.160 uS/cm ELAPSED TIME AT POWER OPERATION: > 10.0 EFPY by F1994 GENERIC LETTER INFORMATION:

I DATE GL 94-03 RESPONSE: Aunust 24.1994 l BASIS FOR ACCElTING JCO: F .s of VT-1 a==^-<aan nerfors d darism W1994 RFO (Cvele 14 RFO)

DATE OF CORE SHROUD INSPECTION /REPAlR

In death UT scheduled for Cycle 15 RFO (F1995) j NRC SERs: (1) SER renardine resnonse to GL 94-03. Januarv 4.1995

(2) Acknowledna aaa! la**ae renardine MSI core shroud la aar*iaa =cana. Aueust 7.1995

!~

l REMARKS: NNECO nerformed enheced VT-1 c=- '- ^1an= of the MSI core =hicaxi d be the Cycle 14 RFO

! (W1994-95). The acaaa of the VT-1 -ena-" lana covered welds Hi-H4 and H6A-H7 from the OD and H3.

i H4. and H5 from the ID. Results of the VT-1 ==--:--*iaan i A a'ad the r,r---- = of only - - r ce=ciriaa in the MSI core ahroud Only 6 i=Aira*ina= were ida=*ified with the loneest haiaa 0. Mans were j- '--- ' = = ^ ner NNECO NCR194 097. Farh.ec,ie. all 3 d cations were less *k-- the iai*ial s. - % .:-

, aanroved for evaluations of the BR-1 core shroud F-dts of the MS1 flaw eviJ_laa instify oner=*iaa of MSl to the Cycle 15 RFO.

l t

B-15 NUREG-1544 s

.- - ~ - - -- - - - . - . . - - . - . - - - . .-

MONTICELLO NUCLEAR GENERATION PLANT DATA SHEET LICENSEE: Northern States Power Company (NSP)

PLANT NAME: Monticello Nuclear Generation Plant (MNGP)

DOCKET NO.: 50-263 LICENSE NO.: DPR-22 BWRVIP CATEGORY GROUPING: Catenorv 'C' SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Rattaai== Dry Dock SHROUD SHELL CONSTRUCTION MATERIAL: A 240 TP304 SS plate CARBON CONTENT RANGE SHELL SECTIONS: 0.043 % C - 0.050% C l SHROUD RING CONSTRUCTION MATERIAL: A 182 F304 SS forrines  !

CARBON CONTENT RANGE SHROUD RINGS: 0.031 % C - 0.056% C WELDING DATA:  !

INITIAL PASS TECHNIQUE:_Ln==>=1 GTAW. =====! SM AW. or automatic S AW

. INITIAL PASS WELD MATERIAL: E308 or ER 308 SUBSEQUENT PASS TECHNIQUE: =====! GTAW. manual SM AW. or automatic SAW  !

SUBSEQUENT PASS WELD MATERIAL: E308 or ER308 OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.299 uS/cm (chlorides 10.6 nob) I ELAPSED TIME AT POWER OPERATION: > 17.8 EFPY by F1994 GENERIC LETTER INFORMATION: ,

DATE GL 94-03 RESPONSE: Aunust 23.1994 1 BASIS FOR ACCEITING JCO: Generic Safety A=====aat. extremely short time to F1994 RFO (Sept.1994) l DATE OF CORE SHROUD INSPECTION / REPAIR: In deoth UT examinations durine F1994 RFO (Sept.1994) 1 NRC SERs: (1) SER renardine resnonse to GL 94-03 and NSP's core shroud inspection results and )

flaw evaluations. January 20.1995  !

REMARKS: In daaek UT/VT-1 a===ia=# ions of the MNGP core shroud durine the Fl994 RFO f%t.15.1994). I UT la==actiaas nerformed with GE OD-Tracker usine 45'S and 60'L transducers. Creeoine wave uced for near j side surface e===inations. ta=naction scope covered UT examiantions of accessible nortions of the H1- H5 welds from the OD (coverane ranned from 32%-54 %). and enhanced VT-1 examinations of the H4. H5. H6. H8. and H9 welds to =>= ale-* the UT ====:-"iaa= ( = 8-15 % coverane). Welds H6-H9 were ia=ccansible to the trackar (Weld H7 was notable to the obstruction from a backins rine). All indications less than 0.25 m (10 in) in leneth. j All 3-Aciaan less th== initial scraaaine criteria li' nit anoroved for evaluation of the BR-1 shroud Operation i iustified. I i

i NUREG-1544 B-16

i

, PEACH BOTTOM UNIT 2 DATA SIIEET l i

i LICENSEE: Philadalakia E!actric Co-av (PECo) '

PLANT NAME: Peach Bottom Akimic Power Station. Unit 2 (PB-2)

DOCKET NO.: 50-277 )

i j LICENSE NO.: DPR-44 _

i BWRVIP CATEGORY GROUPING: Catenoiv "C" l

{ SHROUD FABRICATION DATA:

l SHROUD MANUFACTURER: Rotterd== Dry Dock SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tyne 304 SS clate CARBON CONTENT RANGE SHELL SECTIONS: 0.056 % C - 0.062 % C i SHROUD RING CONSTRUCTION MATERIAL: A 182 F304 SS foreines l CARBON CONTENT RANGE SHROUD RINGS: 0.028% C - 0.035% C WELDING DATA:

INITIAL PASS TECHNIQUE: H7 by GMAW: other welds by SAW INITIAL PASS WELD MATERIAL: H7 usine Allov 82: other welds usine ASTM A371 ER308 l SUBSEQUENT PASS TECHNIQUE: H7 by GMAW: other welds by SAW SUBSEQUENT PASS WELD MATERIAL: H7 usine Allov 82: other welds usira ASTM A317 ER308 OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER l THE FIRST FIVE YEARS OF OPERATION: 0.593 nS/cm '

ELAPSED TIME hT POWER OPERATION: > 11.8 EFPY hv F1994 I

GENERIC LETTER INFORMATION:

DATE GL 9443 RESPONSE: Aunust 24.1994 BASIS FOR ACCEPTING JCO: Any crackine at PB-2 would be to=%1 by evaluations of crackino at PB-3 DATE OF CORE SHROUD INSPECTION / REPAIR: RFO 2R010. Sept./Oct.1994 NRC SERs: (1) SER renardine the PECo resnonse to GL 94-03 for the Peach Bottom Units. Februarv 6.1995 REMARKS: PEco's nerformed UT 3a-tions of the PB-2 core shroud durine RFO 2R010. UT a===iantions were nerformed ~% the GE OD Traakar (45'S. 60'L and ci=-4--e wave tr==- Mrs). UT e===iam' ions covered 30% of weld Hl. 83 %-89 % of welds H2-H5. and 9-10% of welds H6 and H7. An additional 13 % of weld H6 was ' ;ad by " = ' VT-1 = -L 4=. UT av --Nions reva=IaA mma i=Aie=tions at welds Hl. H3. and H4.

naamliaa 0.861 m (33.9 in).1.74 m (68.5 in). and 0.292 m (11.5 in) in !==th. rasaaetively. Extra _-iv minar e=ckian was also da* Mad at welds H5 and H6 (one indication at each weld. both less than 0.13 m l5 inl in lennth).

Flaw ev 1-*ia== ladicada that safety mareins raank 2-=x for the PB-2 shroud would be satisfied for the next cycle.

B-17 NUREG-1544

PEACH BOTTOM UNIT 3 DATA SHEET LICENSEE: Pluladelohia Electric Comnany (PEco)

PLANT NAME: P-h Ra**om Ma ic Power Se=*ian. Unit 3 (PB-3)

DOCKET NO.: 50-278 LICENSE NO.: DPR-56 BWRVIP CATEGORY GROUPING: Catenorv "C"

, SHROUD FABRICATION DATA: )

SHROUD MANUFACTURER: Ra**aal== Dry Dock SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tyne 304 SS ninte CARBON CONTENT RANGE SHELL SECTIONS: 0.050% C - 0.065% C I SHROUD RINO CONSTRUCTION MATERIAL: A 182 F304 SS formines l CARBON CONTENT RANGE SHROUD RINGS: 0.030% C - 0.035% C WELDING DATA:

INITIAL PASS TECHNIQUE: H7 by GMAW: other welds by SAW INITIAL PASS WELD MATERIAL: H7 a ia= Allov 82: a*hae welds unia ASTM A371 ER308 SUBSEQUENT PASS TECHNIQUE: H7 by GMAW: other welds by SAW )

SUBSEQUENT PASS WELD MATERIAL: H7 usine Allov 82: other welds usine ASTM A317 ER308 i OPERATIONAL DATA: '

I AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.695 uS/cm ELAPSED TIME AT PCWER OPERATION: > 11.0 EFPY as of F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aunust 24.1994 BASIS FOR ACCEITING JCO: VT-1 a==' ia== and ev=!a=*in== of the PB-3 iA.cowd iustify cs-eiion of PB3.

DATE OF CORE SHROUD INSPECTION / REPAIR: F '- +i VT-1 -==+ ^iaa= darian RFO 3R9. Oct.1993 NRC SERs: (1) SER rewardine PEco remmae to GL 94-03 for the Peach Bottom Units. Fpruary 6.1995 w -

REMARKS: F '- +j VT-1 au- :--*iaan s fern d durine RFO 3R9. t=+Niaa s~ covered 100% of the acca==ible areas of the H3 and H4 welds from the ID and 100% of the accessible area of the H4 weld from the OD.

" -2 VT-1 == ^ ^!aan were nerfere. on the H6. H7. and H8 welds. In =Adiniaa. one vedical weld k

• L; .- the H3 and H4 wdds was =--- ' The VT-1 e= ==I=="iaas IA=*ified esir- crrLine at weld H3. *+"--

2.67 m (105 in) in leneth. I- ame==ive crackino 'A=*ified at weld H4 (< 0.76 m f 30 inl in leneth). and =laar cranaria- iA-*ified at weld H1 and the vertical weld. The ranule= of PEco's flaw eva!--*ian of the PB-3 core shroud instify onernhon of the PB-3 unit for the evele I- *-e to RFO 3R10 (F1995).

~ NUREG-1544 B-18

FITZPATRICK DATA SHEET LICENSEE: Power Authority of the State of New York (NYPA)

PLANT NAME: James A. FitzPatrick Nuclear Power Plant (FITZ)

DOCKET NO.: 50-333 LICENSE NO.: DPR-59 BWRVIP CATEGORY GROUPING: Catecorv *C" SIIROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Shin Buildine and Dry Dock Co.

SHROUD SHELL CONSTRUCTION MATERIAL: A240 Tvoe 304 SS olates CARBON CONTENT RANGE SHELL SECTIONS: 0.036% C - 0.069% C SHROUD RING CONSTRUCTION MATERIAL: A240 Tvoe 304 SS plates CARBON CONTENT RANGE SHROUD RINGS: 0.056 % C - 0.078 % C WELDING DATA:

INITIAL PASS TECHNIQUE: Fabrication should be similar to that of the Bmnswick Units.

INITIAL PASS WELD MATERIAL: E308 weld wire SUBSEQUENT PASS TECHNIQUE: Fabrication should be similar to that of the Brunswick Units.

, SUBSEQUENT PASS WELD MATERIAL: E308 weld wire OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.631 uS/cm NYPA calculation (0.718 uS/cm by BWRVIP)

ELAPSED TIME AT POWER OPERATION: > 12.8 EFPY by F1994 GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 24.1994 BASIS FOR ACCEPTING JCO: Short duration until W1994-95 RFO: bounded by evaluations of BR-1 DATE OF CORE SHROUD INSPECTION / REPAIR: Repair (modification) W1994-95 RFO NRC SERs: (1) SER rerarding FITZ core shrotd repair. January 5.1995 (2) SER recarding NYPA's response to GL 94-03. February 5.1995 REMARKS: NYPA onted to perform a modification of the FITZ core shroud in lieu of nerformine ccumnaeasive -

core shroud examinations. The tw4 core shroud modification desien involves installation of a series of MPR-desien tie rod assemblies around the circumference of the shroud. The tie rod assemblies are designed to assung the loading of the shroud in the event of a 360' throurh-wall failure of the shroud durine normal werstina. transient and costulated desien basis accident conditions. The NRC approved NYPA's proposed shroud modification design on January 5.1995.

B-19 NUREG-1544

i

?

4

\

. BROWNS FERRY UNIT 1 DATA SHEET l l

LICENSEE: Tennessee Vallev Authority (TVA) l

- PLANT NAME: Browns Ferry Nuclear Plant. Unit 1 (BF-1) I DOCKET NO.: 50-259 LICENSE NO.: DPR-33 BWRVIP CATEGORY GROUPING: Catenorv 'C" l 1

SHROUD FABRICATION DATA: .

SHROUD MANUFACTURER: Rotterdam Dry Dock ~

SHROUD SHELL CONSTRUCTION MATERIAL: A240 Tvoe 304 SS otates (inconel 600 below 117)

CARBON CONTENT RANGE SHELL SECTIONS: 0.030% C -'O.060% C SHROUD RINO CONSTRUCTION MATERIAL: A182 F304 formines CARBON CONTENT RANGE SHROUD RINGS: 0.030% C - 0.060% C WELDING DATA:

INITIAL PASS TECHNIQUE: Techniaues not provided in the response to GL 94-03 INITIAL PASS WELD MATERIAL: E308 or ER308 (INC0182 for H7)

SUBSEQUENT PASS TECHNIQUE: Techniaues not orovided in the response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: E308 or ER308 (INCOl82 for H7)

OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.364 uS/cm ELAPSED TIME AT POWER OPERATION: = 6.5 EFPY GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 23.1994 BASIS FOR ACCEPTING JCO: Reactor is currently in indefinite shutdown and defueled condition DATE OF CORE SHROUD INSPECTION / REPAIR: Deferred NRC SERs: (1) SER renarding resoonse to GL 94-03. January 13.1995 REMARKS: BF-1 is currently in an indefinite shutdown, defueled condition. TVA has indicated that it would perform inwtions of the BF-1 core shroud orior to any reloadine and restart of the BF-1 reactor. No inspections of the BF-1 reactor are needed at this time.

NUREG-1544 B-20

I l

4 4

BROWNS FERRY UNIT 2 DATA SIIEET l LICENSEE: Tennessee Valley Authority (TVA)

PLANT NAME: Browns Ferry Nuclear Plant. Unit 2 (BF-2) I DOCKET NO.: 50-260 l LICENSE NO.: DPR-52 i l

BWRVIP CATEGORY GROUPING: Catecorv "C" l SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Rotterdam Dry Dock l

SHROUD SHELL CONSTRUCTION MATERIAL

A240 Tvoe 304 SS plates (Inconel 600 below H7)

CARBON CONTENT RANGE SHELL SECTIONS: 0.030% C - 0.060% C l SHROUD RING CONSTRUCrlON MATERIAL: A182 F304 forrines l CARBON CONTENT RANGE SHROUD RINGS: 0.030% C - 0.060% C i 1

WELDING DATA:

INITIAL PASS TECHNIQUE: Techniaue not orovided in the response to GL 94-03

{

INITIAL PASS WELD MATERIAL: E308 or ER308 (INCOl82 for H7)

SUBSEQUENT PASS TECHNIQUE: Techniane not orovided in the response to GL 94-03 ]

SUBSEQUENT PASS WELD MATERIAL: E308 or ER308 (INCOl82 for H7)

OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.384 uS/cm ELAPSED TIME AT POWER OPERATION: > 9.0 EFPY by F1994 l

]

GENERIC LETTER INFORMATION: I DATE GL 94-03 RESPONSE: August 23.1994 BASIS FOR ACCEirrlNG JCO: Imoroved material of construction and short time to F1994 RFO I DATE OF CORE SHROUD INSPECTION / REPAIR: Insoected September / October 1994 NRC SERs: (1) SER regardine response to GL 94-03. January 13. 1995 i l

I REMARKS: TVA performed UT insoections of the BF-2 core shroud durine the September / October 1994 RFO.

The UT ia-tions were nerformed mine the GE OD Tracker System. The UT examinations included 45'S. 60'L.

and creenine wave techniaues. fa-tions covered approxinutalv 33% of weld Hl. 61-63% of welds H2-H5 and 2-3% welds H6 and H7. The maiority of the H6 and H7 welds were inaccessible to the tracker. The ia-6ons of the BF-2 shroud reve=!M sinar crackine of the shroud at welds H2. H3. and H5. Total leneths of I A catia== at the H2. H3. and H5 welds were all less than 0.25 m (10.0 in). The results of TVA's flaw evaluation of the BF-2 shroud iustifies operation of the BF-2 unit for the current cycle.

l B-21 NUREG-1544 l

l

l BROWNS FERRY UNIT 3 DATA SHEET LICENSEE: Tennessee Valley Authority (TVA)

PLANT NAME: Browns Ferry Nuclear Plant. Unit 3 (BF-3)

DOCKET NO.: 50-2 %

LICENSE NO.: DPR-68 BWRVIP CATEGORY GROUPING: Catenorv *C" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Rotterdam Dry Dock SHROUD SHELL CONSTRUCTION MATERIAL: A240 Tvoe 304 SS olates (inconel 600 below H7)

CARBON CONTENT RANGE SHELL SECTIONS: 0.030% C - 0.060% C SHROUD RING CONSTRUCTION MATERIAL: /.!82 F304 foreines CARBON CONTENT RANGE SHROUD RINGS: 0.030% C - 0.060% C WELDING DATA:

INITIAL PASS TECHNIQUE: Techniane not orovided in the response to GL 94-03 INITIAL PASS WELD MATERIAL: E308 or ER308 (INCOl82 for H7)

SUBSEQUENT PASS TECHNIQUE: Techniaue not provided in the response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: E308 or ER308 (INCOl82 for H7) l OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER 1 THE FIRST FIVE YEARS OF OPERATION: 0.303 nS/cm ELAPSED TIME AT POWER OPERATION: = 5.0 EFPY GENERIC LETTER INFORMATION:

DATE GL 9403 RESPONSE: Aueust 23.1994 BASIS FOR ACCEIrTING JCO: Reactor is current!v in indefinite defueled condition ._

DATE OF CORE SHROUD INSPECTION / REPAIR: UT and VT-1 inspections durine July 1994 RFO NRC SERs: (1) SER renardine response to GL 94-03. January 13.1995 REMARKS: TVA cerformed UT and VT-1 inspections of the BF-3 core shroud durine the July 1994 RFO. The UT insocctions were performed usine the GE OD Tracker System. The UT examinations included 45'S. 60*L.

and crecoing wave techniaues. The insoections covered aooroximatelv 40-41 % of welds H1 and H5. 68-83%

of welds H2-HA. and 4% of welds H6 and H7. The maiority of the H6 and H7 welds were inaccessible to the tracker. The inscections of the BF-3 shroud revealed some crackine at H5. totaling 2.1 m (82 in)in leneth. Minor crackine of was deteted at shroud welds H1 and H4 (less than 0.05 m l2 inl in total leneth). The results of TVA's fisw evaluation of the BF-3 shroud satisfy ASME safety mareins.

l NUREG-1544 B-22

j VERMONT YANKEE NUCLEAR POWER PLANT DATA SHEET a

s LICENSEE: Vermont Yankee Nuclear Power Corooration (VYNPC)

PLANT NAME: Vermont Yankee Nuclear Power Station (VY) i DOCKET NO.: 50-271 LICENSE NO.: DPR-28 BWRV10 C!.TEGORY GROUPING: Cateeorv *C" i

SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Rotterdam Dry Dock S

SHROUD SHELL CONSTRUCTION MATERIAL: ASTM A240 Tvoe =*=ialess steel olates CARBON CONTENT RANGE SHELL SECTIONS: s 0.070 % C

SHROUD RING CONSTRUCTION MATERIAL: ASTM A182 Grade F304 stainless steel forcines

CARBON CONTENT RANGE SHROUD RINGS

< 0.060% C 4

WELDING DATA:

INITIAL PASS TECHNIQUE: Information not orovided in the resoonse to GL 94-03  !

INITIAL PASS WELD M ATERIAL: Information not provided in the response to GL 944)3 SUBSEQUENT PASS TECIINIQUE: Information not provided in the response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Information not orovided in the response to GL 94-03

]

OPERATIONAL DATA: {

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YLARS OF OPERATION:. 0.286 uS/cm l

ELAPSED TIME AT POWER OPERATION: > 16.9 EFPY by F1994 l

GENERIC LETTER INFORMATION: 1 DATE GL 94-03 RESPONSE: August 17.1994 I BASIS FOR ACCEITING JCO: Generic Safety Assessment and results of RFO No.17 VT-1 e====. (Oct. 93) l DATE OF CORE SHROUD INSPECTION / REPAIR: Cc,...,id.asive UT. April 1995 (RFO No.18)

NRC SERs: (1) SER recardine initial response to GL 94-03. J=an=ry 5.1995 (2) Acknowledeement letter of VY core shroud inspection scope. April 25.1995 (3) SER recardine VY core shroud flaw evaluation. April 27.1995 REMARKS: Cc,w.sid.maive UT ia=aartion of the VY shroud usine new ia=aaction technolony by Babcock and Wilcox Nuclear Technolony (BWNT). BWNT inso-ction tachaology nooroved for use by NRC. UT a===ia=* ions included 45'S. 60*L. and creeping wave UT techniaues. Ramalts of core shroud a===ia=* ions iadirana the or==aare of nienifir==* crackin- at weld H5. Cracks at the H5 welds encompass mooroni-tely 11/12 of weld's circumference Flaw evaluations of H5 weld indicate that the H5 weld will have sufficient ra==iaiae structural li- -* to iustify operation for the next cycle. Evaluations of any crackine discovered at other & C.--d=1 welds bounded by evaluation of cracking at weld H5.

B-23 NUREG-1544

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

SUSQUEHANNA UNIT 1 DATA SHEET LICENSEE: Pennsviv- l- l'ower & Heht C<=====v (PP&L) '

FLANT NAME: Susquehanra steam Electric Station. Unit 1 (SSES-1)

DOCKET NO.: 50-387 LICENSE NO.: NPF-14 BWRVIP CATEGORY GROUPING: Cateeorv "B" l i

, SHROUD FABRICATION DATA:

]

SHROUD MANUFACTURER: Chicano Bridae and Iron Works 1 SHROUD SHELL CONSTRUCTION MATERIAL: SA 240 TP 304L hot rolled plate CARBON CONTENT RANGE SHELL SECTIONS: 0.014% C - 0.027% C SHROUD RING CONSTRUCTION MATERIAL: SA 240 TP 304L hot rolled plate CARBON CONTENT RANGE SHROUD RINGS: 0.025% C - 0.026% C i]

WELDING DATA: ,

INITIAL PASS TECHNIQUE: SMAW for H1-H6 and H8-H9 welds: GTAW for H7 weld [

INITIAL PASS WELD MATERIAL: SFA 5.4 E308 for SM AW: SFA 5.14 ERNiCr-3 for GTAW t SUBSEQUENT PASS TECHNIQUE: SAW for Hl-H3 and H6a-H6b welds. SMAW for H4-H5. H7-H9  ;

SUBSEQUENT PASS WELD MATERIAL:SFA E'.iOS most welds: SPA 5.9 ER308L for Hl: ERNiCrFe-3 for H7 1 OPERATIONAL DATA:  !

AVERAGE CONDUCrlVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.205 nS/cm ELAPSED TIME AT POWER OPERATION: ~ 9 EFPY at the time of RFO No. 8RIO .

, GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aueust 24.1994 BASIS FOR ACCEPTING JCO:Cateeorv "B" criteria and results of limited VT-Is at RFO No. 7RIO (Fl993) f DATE OF CORE SHROUD INSPECTION / REPAIR: I iraisad UT === ia-*ians durine Sorine 1995 RFO (8RIO) i NRC SERs: (1) SER aae. "== ra-aa- to GL 94-03. March 23,1995.

(2) Aciraawlad-* ta**ar of SSES-1 lamaaetion scana. Anril 10. 1995.* ,

(3) SER reenrdine the SSES 1 core shroud insocction results and flaw evaluation. May 3.1995. I r

REMARKS: 1 iraital la aareina of the SSES-1 core shroud durine RFO No. 8RIO. Scone in =ecaedaare with the

. ract - aaada*iaa= of the BWRVIP for F' Tsv *B" core shrouda. Ia aactians included UT of H3. H4. H5. H6a. ,

H6b. and H7 welds. and --"M VT-1 of the H8 and H9 walds R- ":s of the UT !==--:^ aan reveelad a i

- mi.aific-e au-ha, of flaw i=dia=*ina= at welds H2. H4. H5. and H6b welds tas=Iline 13.8 m (54.4 in). 4.76 m

' (187.5 in). 4.80 m (189.9 in). and 1.66 m (65.3 in)in leneth at each weld. respectively. Minor amount of crackine j 3130 detected welds H1 and H6A (less *kaa 0.13 m 15 ini total at ==ch weld). F-ui st of Hoections of other welds  ;

were manative for i d:-=*iaa=. Flaw evalu=*inam piaaart operation for the next operatine cycle. l t

i NUREG-1544 B-24

SUSQUEHANNA UNIT 2 DATA SHEET 1

LICENSEE: Pennsylvania Power & Licht Company (PP&L)

PLANT NAME: Susauehanna Steam Electric Station. Unit 2 (SSES-2)

DOCKET NO.: 50-388

. LICENSE NO.: NPF-22 l l

4 BWRVIP CATEGORY GROUPING: Categorv "B" SHROUD FABRICATION DATA:

, SHROUD MANUFACTURER: Chicano Bridae and iron Works SHROUD SHELL CONSTRUCTION MATERIAL: SA 240 TP 304L hot rolled clate I

CARBON CONTENT RANGE SHELL SECTIONS: 0.014 % C - 0.025 % C SHROUD RING CONSTRUCTION MATERIAL: SA 240 TP 304L hot rolled plate CARBON CONTENT RANGE SHROUD RINGS: 0.025% C - 0.027% C 1

WELDING DATA:

1 INITIAL PASS TECHNIQUE: SMAW for H1-H6 and HB-H9 welds: GTAW for H7 weld INITIAL PASS WELD MATERIAL: SFA 5.4 E308 for SMAW: SFA 5.14 ERNiCr-3 for GTAW SUBSEQUENT PASS TECHNIQUE: SAW for H1-H3 and H6a-H6b welds. SMAW for H4-H5. H7-H9 SUBSEQUENT PASS WELD MATERIAL:SFA E308 most welds: SFA 5.9 ER308L for Hl: ERNiCrFe-3 for H7 OPERATIONAL DATA:

I AVERAGE CONDUCTIVITY VALUE OVER

, THE FIRST FIVE YEARS OF OPERATION: 0.198 uS/cm j ELAPSED TIME AT POWER OPERATION: 6.2 EFPY at RFO No. 6RIO: ~ 8.0 EPFY at RFO No. 7RIO

GENERIC LETTER INFORMATION:

) DATE GL 94-03 RESPONSE: Aueust 24.1994 BASIS FOR ACCEPTING JCO:Cateeory "B" criteria l DATE OF CORE SHROUD INSPECTION / REPAIR: Scheduled for U2 RFO No. 7RIO

. NRC SERs: (1) SER renardine PP&L's resoonse to GL 94-03 for SSES-1/SSES2. March 23.1995.

REMARKS: No limited ex==iaasions nerformed durine U2 RFO No. 6RIO since unit had not ooerated for more thaa 8 EPFY.1 imited eraminations sch-bled for U2 RFO No. 7RIO in Seotember/ October 1995.

B-25 NUREG-1544

LASALLE UNIT 1 DATA SHEET LICENSEE: Commonwealth Edison comoany PLANT NAME: lasalle Unit 1 (LA-1)

DOCKET NO.: 50-373 LICENSE NO.: NPF-11 BWRVIP CATEGORY GROUPING: Cateeorv *B" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Shin Buildine and Dry Dock Co.

SHROUD SHELL CONSTRUCTION MATERIAL: SA 240 Tvoe 304L Stainless Steel CARBON CONTENT RANGE SHELL SECTIONS: 0.019% C - 0.028% C  ;

SHROUD RING CONSTRUCTION MATERIAL: SA 240 Tvoe 304L Stainless Steel i CARBON CONTENT RANGE SHROUD RINGS: 0.021 % C - 0.024% C ,

WELDING DATA:

INITIAL PASS TECHNIQUE: Not orovided in the resoonse to GL 94-03 INITIAL PASS WELD M ATERIAL: Not orovided in the response to GL 94-03 SUBSEQUENT PASS TECHNIQUE: Not orovided in the response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Not orovided in the resgong to GL 94-03 OPERATIONAL DATA:

AVERAGE CONDUCTIVTTY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.272 nS/cm ELAPSED TIME AT POWER OPERATION: As of February 1994. 8.04 years at oower GENERIC LETTER INFORMATION:

DATE GL 9443 RESPONSE: Aueust 23.1994 BASIS FOR ACCEI' TING JCO: Inspection results of April 1994 core shroud inspections.

DATE OF CORE SHROUD INSPECTION / REPAIR: Insoection. April 20-May 2.1994.

NRC SERs: (1) SER renardine the Comed response to Generic Letter for LA-1/LA-2. February 16. 1995.

REMARKS: Enhanced VT-1 insoection of the LA-1 shroud usine ETV-1250 black and white video camera with

" twin 50' tvoe liehtine, insoections covered accessible nortions of welds Hi-H8 from the OD. and portions of the H3. H4. and H5 welds that were accessible from the ID throueh open cell locations durine refueline. No indications detected durine the examinations.

1 i

NUREG-1544 B-26  :

i i

1 l

l LASALLE UNIT 2 DATA SHEET i 1

LICENSEE: Cr -- = a-1*h EAi== C.===v (Co=W)

PLANT NAME: I - ile Unit 2 (LA-2)

DOCKET NO.: 50-374 LICENSE NO.: NPF-18 BWRVIP CATEGORY GROUPING: Catenorv *B" I SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Eabuildina and Dry Dock Co.

SHROUD SHELL CONSTRUCTION MATERIAL: SA 240 Tyne 304L 9 ial==s Steel CARBON CONTENT RANGE SHELL SECTIONS: 0.018 C - 0.024% C  ;

SHROUD RING CONSTRUCTION MATERIAL: SA 240 Tvoe 304L 9-1=laas Steel  !

CARBON CONTENT RANGE SHROUD RINGS: 0.022% C - 0.028% C l WELDING DATA:

INITIAL PASS TECHNIQUE: Not provided in the response to GL 94-03 INITIAL PASS WELD MATERIAL: Not orovided in the renoonse to GL 94-03 SUBSEQUENT PASS TECHNIQUE: Not orovided in the resnonse to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Not provioed in the resnonse to GL 94-03 .

1 l

OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER l THE FIRST FIVE YEARS OF OPERATION: 0.272 uS/cm

~

ELAPSED TIME AT POWER OPERATION: 8.53 EFPY oroiected to the February 1995 RFO GENERIC LE' ITER INFORMATION:

DATE GL 94-03 RESPONSE: Aucunt 23.1994 BASIS FOR ACCEPTING JCO: Pa=1ts of the Anril 1994 LA 1 core shroud iamaaetions bm Al== for LA-2.

DATE OF CORE SHROUD INSPECTION / REPAIR: Insnection. March 20-Aoril 10.1995.

NRC SERs: (1) SER renardia= the Ca=W raarw=aa to Gaaaaic La**er for LA-1/LA-2. February 16. 1995.

(2) Achaawledna-* LMer of LA-2 core shroud i==aareiaa acnaa. Februarv 6.1995.

REMARKS: UT I==+11aa of the LA-2 hi d wd- r- I ==!== the GE O.D. Tids. ind=" - 45*S. 60'l;.

and craaaaaa wave UT *ack-h. Ia-tians covered = era ==ikla nortinaa of welds H3-H6 and weld H8 from the OD. No iaAiraeinas daeactad durinn the a=- :--*iaan-B-27 NUREG-1544

A w. m, w l. ,F MG -6 n. A a . --% d4-- w- km--e '

$s ~ - + - - e. +s5*- 4 A 'Mo e,. 4- ----*w- -

GRAND GULF UNIT 1 DATA SHEET LICENSEE: Enterav Operations. Incoroorsted (EOl)

PLANT NAME: Grand Gulf Unit 1 (GG-1)

DOCKET NO.: 50-416 LICENSE NO.: NPF-29 BWRVIP CATEGORY GROUPING: Catenorv *B" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Binoh== Willi ==a'te SHROUD SHELL CONSTRUCTION MATERIAL: SA 240 Tyne 304L hot Rolled Plate CARBON CONTENT RANGE SHELL SECTIONS: 0.015% C - 0.019% C SIIROUD RING CONSTRUCTION MATERIAL: SA 240 Tyne 304L hot Rolled Plate CARBON CONTENT RANGE SHROUD RINGS: 0.011 % C - 0.016% C WELDING DATA:

INITIAL PASS TECHNIQUE: H1-H6B welds. SAW INITIAL PASS WELD MATERIAL: ER308L filler metal SUBSEQUENT PASS TECHNIQUE: H1-H6B welds. SAW SUBSEQUENT PASS WELD MATERIAL: ER308L filler metal OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.0235 aS/cm (0.222 nS/cm full life to date) ,

ELAPSED TIME AT POWER OPERATION: = 8.6 EFPY as of April 1995 RFO (No. RFO7)

GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aueust 19. 1994 BASIS FOR ACCEPTING JCO: Catenorv *B' criteria DATE OF CORE SHROUD INSPECTION / REPAIR: Limited inspections performed April 1995 RFO (No. RFO7)

NRC SERs: (1) SER renardine the response to GL 94-03 for GG-1. March 29.1995 REMARKS: Limited UT inspections of the GG-1 core shroud performed in accordance with the BWRVIP ruidelines for Catenorv *B" core shrouds. Inspection scone covered H3. H4. H6A. and H7 welds. Creenine wava

- techniaue used as a supplement for examinine the OD surface. No evidence of crackine detected as a result of the eraminatiOns.

NUREG-1544 D-28

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LIMERICK UNIT 1 DATA SIIEET 1

l LICENSEE: ~ Philadelphia Electric Cama =av (PEco) j PLANT NAME: Limerick Unit 1 (LIM 1) i DOCKET NO.: 50-352 '

' LICENSE NO.: NPF-39 EWRVIP CATEGORY GROUPING: Catenorv *B" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Shin Bui1Aiao and Dry Dock

)

SHROUD SHELL CONSTRUCTION MATERIAL:. A 240 Tvoe 304L Stainless Steel CARBON CONTENT RANGE SHELL SECTIONS: 0.018% C - 0.024% C SHROUD RING CONSTRUCTION MATERIAL: A 240 Tyne 304L Stainless Steel CARBON CONTENT RANGE SHROUD RINGS: 0.024% C - 0.026% C WELDING DATA:

~ INITIAL PASS TECHNIQUE: Not orovided in the response to GL 94-03 INITIAL PASS WELD MATERIAL: Not provided in the resnonse to GL 04-03 SUBSEQUENT PASS TECHNIQUE: Not provided in the response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Not orovided in the response to GL 9443 OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.15 uS/cm ELAPSED TIME AT POWER OPERATION: As of Aueust 24.1994. 6.4 EFPY GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aunust 24.1994 BASIS FOR ACCEPTING JCO: Cateeorv *B' criteria: averaee RCS conductivity. shroud construction material DATE OF CORE SHROUD INSPECTION / REPAIR: Limited inspection scheduled for RFO 1R06. January 1996 NRC SERs: (1) SER renardine the PEco response to GL 94-03 for LIM-l. March 7.1995 l

REMARKS: '

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B-29 NUREG-1544 i

y. . . - _ , . ..~ .- -. n - .- .s - - a- . -- n..- --.

l HOPE CREEK UNIT 1 DATA SHEET LICENSEE: Public Service Electric and Gas Comoany (PSE&G)

PLANT NAME: Hone Creek Station Unit 1 (HC-1) i DOCKET NO.: 50-354 l LICENSE NO.: NPF-57 I BWRVIP CATEGORY GROUPING: Cateeory "A" SHROUD FABRICATION DATA: I SHROUD MANUFACTURER: Rotterdam Dry Dock I SHROUD SHELL CONSTRUCTION MATERIAL: ASTM A-240 Tvoc 304L Stainless Steel Plates CARBON CONTENT RANGE SHELL SECTIONS: 0.008% C - 0.025% C SHROUD RING CONSTRUCTION MATERIAL: SA 102 Tvoe 304L Stainless Steel Foreines 4 CARBON CONTENT RANGE SHROUD RINGS: 0.021 %C - 0.030% C l WELDING DATA:

INITIAL PASS TECHNIQUE: Not provided in the response to GL 94-03 INITIAL PASS WELD MATERIAL: Not provided in the resoonse to GL 94-03 SUBSEQUENT PASS TECHNIQUE: Not provided in the response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Not provided in the resoonse to GL 94-03 OPERATIONAL DATA:  ;

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.155 uS/cm ELAPSED TIME AT POWER OPERATION: 7.7 EFPY by the 6th RFO GENERIC LETTER INFORMATION:

l DATE GL 94-03 RESPONSE: Aueust 24.1994 BASIS FOR ACCEPTING JCO: Cateeory "A" criteria  ;

DATE OF CORE SHROUD INSPECTION / REPAIR: Deferred to the March 1997 RFO NRC SERs: (1) SER reenrdine the PSE&G response to GL 94-03 for 11C-1 (issue date was October 10. 1995).

1 1

REMARKS: The SER for Hooe Creek was issued on October 10.1995. This date is outside the time frame scope of this NUREG. However. the date is listed in this data sheet to indicate that an SER was issued to PSE&G regardine its response to GL 94-03.

NUREG-1544 B-30

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. NINE MILE POINT UNIT 2 DATA SHEET l

1 LICENSEE:__ Niagara Mohawk Power Corporation (NMPC)

PLANT NAME: Nine Mile Point Unit 2 (NMP-2)

DOCKET NO.: 50-410 LICENSE NO.: NPF-69 1

BWRVIP CATEGORY GROUPING: Cateeorv "A" l

SliROUD FABRICATION DATA:  !

SHROUD MANUFACTURER: Sun Shin Buildina and Dry Dock SIIROUD SHELL CONSTRUCTION MATERIAL: SA 240 Tvoe 304L Plate I CARBON CONTENT RANGE SHELL SECTIONS: 0.017 % C - 0.030% C SHROUD RING CONSTRUCTION MATERIAL: S A 240 Tvoc 304L Plate CARBON CONTENT RANGE SHROUD RINGS: 0.014 % C - 0.021 % C WELDING DATA:

INITIAL PASS TECHNIQUE: SAW for cire. and vert. welds: SM AW for flance serments {

INITIAL PASS WELD MATERIAL:_ ASME SFA-5.9 ER308 (SAW) 1 SUBSEQUENT PASS TECHNIQUE: SAW for cire. and vert, welds: SM AW for flance seements l SUBSEQUENT PASS WELD MATERIAL: ASME SFA-5.9 ER308 (SAW) 0*ERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER l

THE FIRST FIVE YEARS OF OPERATION

0.0129 uS/cm l ELAPSED TIME AT POWER OPERATION: 8 EFPY oroiected in Scotember 1998 assumine 18 month cycles  !

] )

GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 23.1994 j

BASIS FOR ACCEPTING JCO: Catenorv 'A' criteria 1 DATE OF CORE SHROUD INSPECTION / REPAIR: Deferred NRC SERs: SER regarding response to GL 94-03. Feb. 2.1995.

t RE51 ARKS: Limited VT-1 of NMP-2 core shroud performed during the October 1993 RFO (RF-03). No indications were evident as a result of the examinations.

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l B-31 NUREG-1544

l RIVER BEND UNIT 1 DATA SIIEET LICENSEE: Enterev Operations. Incorporated (EOI)

PLANT NAME: River Bend Unit 1 (RVR-1)

DOCKET NO : 50-458 LICENSE NO.: NPF-47 BWRVIP CATEGORY GROUPING: Category " A"

~

l SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Shio Buildine and Dry Dock SHROUD SHELL CONSTRUCTION MATERIAL: SA 240 Tvoe 304L Stainless Steel CARBON CONTENT RANGE SHELL SECTIONS: 0.022 % C - 0.025 % C SHROUD RING CONSTRUCTION MATERIAL: SA 240 Tvoe 304L Stainless Steel CARBON CONTENT RANGE SHROUD RINGS: 0.018% C - 0.029% C WELDING DATA:

INITIAL PASS TECHNIQUE: Automated SAW INITIAL PASS WELD MATERIAL: ER-308L Filler SUBSEQUENT PASS TECHNIQUE: Automated SAW SUBSEQUENT PASS WELD MATERIAL: ER-308L FilleL.

OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER  !

THE FIRST FIVE YEARS OF OPERATION: 0.160 uS/cm ELAPSED TIME AT POWER OPERATION: Proiected to be 7.3 EFPY at RFO No. 6 (Sentember 1995)

GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 24.1994 BASIS FOR ACCEPTING JCO: Category "A" criteria DATE OF CORE SHROUD INSPECTION / REPAIR: Deferred in accordance with Cateeory "A" ruidelines i NRC SERs: SER reeardine response to GL 94-03. February 3.1995 l

REMARKS:

I NUREO-1544 B-32

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FERMI UNIT 2 DATA SHEET j

$ l i LICENSEE: Detroit Edison Comnany (DECO)  !

j PLANT NAME: Fermi Unit 2 (FRM-2) i DOCKET NO.: 50-341 LICENSE NO.: NPF-43 f j e

BWRVIP CATEGORY GROUPING: Catenorv "A" i 1

j {

SHROUD FABRICATION DATA j SHROUD MANUFACTURER: Core Shroud: Rotterd== Dry Dock: Core Support Plate: Combustion Ene.

SHROUD SHELL CONSTRUCTION MATERIAL: 304L rolled plate i j CARBON CONTENT RANGE SHELL SECTIONS: 0.013% C - 0.029% C

! SHROUD RING CONSTRUCTION MATERIAL: 304L forned rines CARBON CONTENT RANGE SHROUD RINGS: 0.020% C - 0.035 % C l I WELDING DATA: .

! INITIAL PASS TECHNIQUE: Not provided in the response to GL 94-03 i

, INITIAL PASS WELD MATERIAL: ER308L 1 SUBSEQUENT PASS TECHNIQUE: Not provided in the response to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: ER308L i OPERATIONAL DATA:

! AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: Better than the norm for the industry: value not provided l ELAPSED TIME AT POWER OPERATION: As of December 25.1993. 4.36 EFPY GENERIC LETTER INFORMATION:

DATE OL 94-03 RESPONSE: Aunust 24.1994 BASIS FOR ACCEPTINO JCO: Cateeory "A" criteria. and results of previous shroud examinations.

DATE OF CORE SHROUD INSPECrlON/ REPAIR: Enhanced VT-1 in accordance with SIL 054. April 1991 NRC SERs: SER renardina Response to GL 94-03. January 24.1995 f

l REMARKS: M--W VT-1 e===iaa' ions nerformed at recom.==.dation of GE after event at KKM.

Examinations included VT-1 of accessible nortions of welds H1-H7 from the OD. and VT-1 of H2. H3. and H4 from the ID. throuah accessible nortions of peripheral fuels cells with control rod blades withdrawn. VT-1 exams nerformed a :== a color ea-a sva**= and a S-VHS hinh resolution video recordine system. ca_aable of resolvine 3 a 1 mil wire (calibration standard). No flaw indications detected as a result of the VT-1 examinations.

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B-33 NUREG-1544 4

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._. _ _ . . _ . . . . . _ . - _ . . _ . . . . _ -_ _ _. _ . . . _ ... _ __. _ _ _.._ _ _ . _m._. _

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LIMERICK UNIT 2 DATA SHEET 1

LICENSEE: Ph141= F1-tdc C-av (PEco)

PLANT NAME: 1 IHek Unit 2 (LIM-2)  !

DOCKET NO.: 50-353 LICENSE NO.: NPF-85 BWRVIP CATEGORY GROUPING: Catenorv 'A' SHROUD FABRICATION DATA:  ;

SHROUD MANUFACTURER: Not orovided in the r~-- to GL 94-03 SHROUD SHELL CONSTRUCTION MATERIAL: A 240 Tyne 304L S* lat=== S**! ,

CARBON CONTENT RANGE SHELL SECTIONS: 0.018% C - 0.024% C SHROUD RING CONSTRUCTION MATERIAL: A 240 Tvom 304L Se=3 1== Steel ,

CARBON CONTENT RANGE SHROUD RINGS: 0.024% C - 0.026% C r

WELDING DATA:

INITIAL PASS TECHNIQUE: Not orovided in the resnonse GL 94-03 l INITIAL PASS WELD MATERIAL: Not provided in the resnonse GL 94-03  !

SUBSEQUENT PASS TECHNIQUE: Not orovided in the response to GL 94-03 )

SUBSEQUENT PASS WELD MATERIAL: Not provided in the response to GL 94-03 l OPERATIONAL DATA: _ _

AVERAGE CONDUCTIVITY VALUE OVER  !

.THE FIRST FIVE YEARS OF OPERATION: 0.123 nS/cm ELAPSED TIME AT POWER OPERATION: As of Aueust 24.1994. 3.6 EFPY GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: Aunust 24.1994 BASIS FOR ACCEITING JCO: Cateeorv "A" criteria DATE OF CORE SHROUD INSPECTION / REPAIR: Deferred in accordance with Catenorv 'A' Guidehnes I NRC SERs: SER aae.-dhis response to GL 94-03. March 13.1995 1

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REMARKS:

I NUREG-1544 B-34

1 CLINTON DATA SIIEET LICENSEE: 11tinois Power Comanav (IPC)

PLANT NAME: Clinton Power Station (CPS)

DOCKET NO.: 50-461 LICENSE NO.: NPF-62 BWRVIP CATEGORY GROUPING: Category "A" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Sun Shio Buildine and Dry Dock (Shroud Support - CBIW)

SHROUD SHELL CONSTRUCTION MATERIAL: SA 240 Tvoe 304L CARBON CONTENT RANGE SHELL SECTIONS: 0.018 % C - 0.021 % C SHROUD RING CONSTRUCTION MATERIAL: SA 240 Tvoe 304L CARBON CONTENT RANGE SHROUD RINGS: 0.024 % C - 0.026% C WELDING DATA:

INITIAL PASS TECHNIQUE: Not orovided in the response to GL 94-03 INITIAL PASS WELD MATERIAL: Either ER308 Filler or ER308L Filler SUBSEQUENT PASS TECHNIQUE: Not orovided in resoonse to GL 94-03 SUBSEQUENT PASS WELD MATERIAL: Either ER308 Filler or ER308L Filler i OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.188 uS/cm ELAPSED TIME AT POWER OPERATION: 5.4 EFPY GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 24. 1994 BASIS FOR ACCEPTING JCO: Category "A" insoection criteria DATE OF CORE SHROUD INSPECTION / REPAIR: Deferred accordine the Category "A" ruidelines NRC SERs: SER renardine response to GL 94-03. Feb.10.1995 REMARKS: (1) Ave. sulfate for first five years: 3.40 oob. (2) Ave. chloride for first five years: 2.10 onb.-

(3) Ave. Chromate for first five years: 34.2 nob.

B-35 NUREG-1544 1

1 PERRY UNIT 1 DATA SHEET LICENSEE: Centerior Enercy. Inc. (CEI) 1 PLANT NAME: Perry Nuclear Power Plant (PRY)

DOCKET NO.: 50-440 )

LICENSE NO.: NPF-58 '

4 BWRVIP CATEGORY GROUPING: Catecorv " A" SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Bineham Williamette SHROUD SIIELL CONSTRUCTION MATERIAL: SA 240 Tyne 304L Stainless Steel CARBON CONTENT RANGE SHELL SECTIONS: 0.02I % C Max.

SilROUD RING CONSTRUCTION MATERIAL: SA 240 TvDe 304L Stainless Steel 1

CARBON CONTENT RANGE SHROUD RINGS: 0.016% C Max.

WELDING DATA:

INITIAL PASS TECHNIQUE: SAW l INITIAL PASS WELD MATERIAL: E308L Filler SUBSEQUENT PASS TECHNIQUE: SAW SUBSEQUENT PASS WELD MATERIAL: E308L Filler OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER THE FIRST FIVE YEARS OF OPERATION: 0.20 uS/Cm (0.13 uS/cm last two cycles)

ELAPSED TIME AT POWER OPERATION: As of August 24.1994. 4.1 EFPY (4 Cycles of operation)

GENERIC LETTER INFORMATION:

DATE GL 94-03 RESPONSE: August 24.1994 BASIS FOR ACCEPTING JCO: Catecorv " A' plant criteria DATE OF CORE SHROUD INSPECTION / REPAIR: Deferred NRC SERs: GL 94-03 Response SER. Febmary 10.1995 1

1 l

1 REMARKS: (1) Limited examinations of the H3 and H4 welds durine RFO No. 4. No evidence of crackine.

Additional insnections deferred until 8 EFPY has been surnassed.

NUREG-1544 B-36

i 4

WNP-2 DATA SIIEET I

LICENSEE:- Washincton Public Power Sunniv System (WPPSS)

PLANT NAME: Washincton Nuclear Plant. Unit 2 (WNP-2) ,

DOCKET NO.: 50-397 , ,

t LICENSE NO.: N PF-21 ,

BWRVIP CATEGORY GROUPING: Categorv "A" _

SHROUD FABRICATION DATA:

SHROUD MANUFACTURER: Chicano Bridce and Iron Works  !

STIROUD SHELL CONSTRUCTION MATERIAL: SA 240 Type 304L Stainless Steel i

CARBON CONTENT R ANGE SHELL SECTIONS:Shell & Rine carbon contents rance: 0.010% C-0.024% C l SHROUD RING CONSTRUCTION MATERIAL: SA 240 Type 304L Stainless Steel  !

CARBON CONTENT RANGE SIIROUD RINGS: Shell & Rine carbon contents rance: 0.010% C-0.024% C WELDING DATA:

INITIAL PASS TECilNIQUE: Notprovided in the response to GL 94-03 INITIAL PASS WELD M ATERI AL: Not provided in the response to GL 94-03 SUBSEQUENT PASS TI CilNIQUE: Not provided in the response to GL 94-03 SUBSEQUENT PA5S WLLD MATERIAL: Not provided in the response to GL 94-03 OPERATIONAL DATA:

AVERAGE CONDUCTIVITY VALUE OVER TIIE FIRST FIVE YEARS OF OPERATION: 0.242 uS/cm (0.175 uS/cm over last five cycles)

ELAPSED TIME AT POWER OPERATION: As of the April 1994 RFO. 5.8 EFPY GENERIC LETTER INFORMATION:

DATE GL 94 03 RESPONSE: Aucust 24.1994 BASIS FOR ACCEf' TING JCO: Catevory "A" plant criteria.

DATE OF CORE SHROUD INSPECTION / REPAIR: Defe. red NRC SERs: SER rerardine response to GL 94-03. May 8.1995 REMARKS: (1) Limited Examination of H3 and H4 welds durine April 1994 RFO. No indications of IGSCC.

(2) Averane shroud Phosphorous Content: 0.020% P.

(3) Averane Shroud Sulfur Content: 0.014 % S.

B-37 NUREG-1544

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APPENDIX C LIST OF BWR UTILITIES AND REACTORS Abbreviation Boston Edison Company . . . . . . . . . . . ................................. BECo l

..... Pilgrim Nuclear Power Station . . . . ......................... PNPS Carolina Power & Light Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . CP&L 1 1

. . . . . Brunswick Steam Electric Plant Unit 1. . . . . . . . . . . . . . . . . . . . . . . . . . . BR-1

. . . . . Brunswick Steam Electric Plant Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . BR-2 1

1 Centerior Energy, lacorporated . . . . . . . . . . . . . . . . . . . . . . . . . . . . ............ CEI

. . . . . Perry Nuclear Power Station . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . PRY Commonwealth Edison Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Comed l . . . . . D resden Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . DR-2 I l . . . . . D resden Unit 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . DR3 I

. . . . . Quad Cities Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . QC-1 l

. . . . . Quad Cities U nit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . QC-2 l . . . . . Lasalle Unit ! . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . LA 1

. . . . . f maalle Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . LA-2 Detroit Edison company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Deco l

i

. . . . . Fermi Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . FRM-2 Entergy Operation, Incorporated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . EOI l

i

. . . . . G rand G ul f Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . GG-1 l . . . . . River Bend Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . RVR-1 t

1 l General Public Utilities . . . . . . . . . . . . . . . . . . . . . . . . ..................... GPU i

l . . . . . Oyster Creek Nuclear Generation Station . . . . . . . . . . . . . . . . . . . . . . . . . . OCNGS 4

4 C-1 NUREG-1544 i

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Abbreviation Georgia Power Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . GPC )

I

. . . . . Edwin 1. Hatch Uni t 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . HAT-1 i

. . . . . Edwin 1. Hatch Unit 2 . . . . . . . . . . . . . . . . . ................... HAT-2

' Illinois Power Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IPC i

. . . . . Clinton Power Station . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . CPS i

IES Utilities, Incorporated . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IES 8

. . . . . Duane Arnold Energy Center . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . DAEC ,.

I r

l Nebraska Public Power District . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

NPPD 1

. . . . . Cooper Nuclear Station . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . CNS Niagara Mohawk Power Corporation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . NMPC

. . . . . Nine M ile Point Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . NMP1 l

. . . . . Nine Mile Point Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . N M P-2 I

I t

i Northeast Nuclear Energy Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . NNECo

+

l

. . . . . M illstone Uni t ! . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . M-1 Northern States Power Company . . . . . . . . . . . . . . . . . . . . . . ................. NSP

. . . . . Monticello Nuclear Generation Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . MNGP Pennsylvania Power & Light Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . PP&L 1

. . . . . Susquehanna Steam Electric Station, Unit 1. . . . . . . . . . . . . . . . . . . . . . . . . SSES-1

. . . . . Susquehanna Steam Electric Station, Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . SSES-2 j Philadelphia Electric Company . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . PEco

)

. . . . . Li merick Unit 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . LIM-1  ;

. . . . . Li merick Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . LIM-2

. . . . . Peach Bottom Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . PB 2

. . . . . Peach Bottom Uni' 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . PB-3 NUREG-1544 C-2

., re , ~ .

i l

Abbreviation Power Authority of the State of New York . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . NYPA

. . . . . James A. Fiti. Patrick Nuclear Pcwer Plant . . . . . . . . . . . . . . . . . . . . . . . . . FITZ Public Services Electric and Gas Conpany . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . PSE&G

. . . . . Hope Creek Station Unit I . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . HC-1 Tennessee Valley Authority . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . TVA

. . . . . Browns Ferry Unit i ...................................... BF-1

. . . . . B rowns Ferry Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . BF-2

... . Browns Ferry Unit 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . BF-3 j J

Vermont Yankee Nuclear Power Corporation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . VYNPC j i

. . . . . Vermont Yankee Nuclear Power Station . . . . . . . . . . . . . . . . . . . . . . . . . . VY Washington Public Power Supply System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . WPPSS

. . . . . Washington Nuclear Plant Unit 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . WNP-2

)

e C-3 NUREG-1544

APPENDIX D ABBREVIATIONS AND NOMENCLATURE 10 CFR Title 10, Code of Federal Reculations GTAW gas tungsten are welding AHC Access Hole Cover HAZ heat affected zone ASME American Society of Mechanical Engineers ID inner diameter ASTM American Society for Testing and Materials IGSCC intergranular stress corrosion cracking BWNT Babcock and Wilcox Nuclear Technology IN information notice BWR boiling water reactor JCO justification for continued operation BWROG Boiling Water Reactor Owners Group JPHDD jet pump hold down beam

BWRVIP BWR Vessel and Internals Project LEFM linear elastic fracture mechanics cire. circumferential LLA limit load analysis CS core spray system LOCA loss of coolant accident ECCS emergency core cooling systems MPR MPR Associates ECP electrochemical potential MSLB main steam line break

. EDM electrodischarge machined NCR Nonconformance Report EFPY effective full-power years NDE nondestructive examination EOC end of cycle NRC Nuclear Regulatory Commission i EOL end oflife OD outer diameter EPFM elastic plastic fracture mechanics PRA probabilistic risk assessment EPRI Electric Power Research Institute RAI request for additional information ET eddy current testing RCS reactor coolant system F19xx Fall, Year 19xx RFO refueling outage GE General Electric Company RICSIL Rapid Information Communication Services information Letter OL generic letter GMAW gas metal arc welding RLB recirculation line break D-1 NUREG-1544

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

i l

l l

RPV reactor pressurc vessel UI Unit i S19xx Summer, Year 19xx U2 Unit 2 l

t SAW submerged arc welding U3 Unit 3 SER safety evaluation report UT ultrasonic testing

! SIL Services information Letter VT visual testing SLCS standby liqi Id control system vert. vertical l SMAW shielded metal are welding W19xx Winter, Year 19xx l

SP19xx Spring, Year 19sx W/SP19xx Late Winter, Early Spring, Year 19xx SS stainten stect

)

Commnniv Used Scientific Units i

in length in when pS/cm microSiemens per centimeter, unit of electrical I conductivity (this is equivalent to a unit in in/hr velocity or gniath rate in inches per hour mhos/cm, micrombos per centimeter)

I m length in meters V volt, unit of electrochemical potential (ECP) m/s velocity or gnemth rate in meters per second E

a 3

l NUREG-1544 D-2

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APPENDIX E LIST OF STAFF CONTRIBUTORS TO NUREG-1544 This report was compiled by Mr. James Medoff of the Materials and Chemical Engineering Branch in the Office of Nuclear Reactor Regulation, Division of Engineering. Other contributors to this report and to the staff's efforts regarding the issue of IGSCC in BWR internal components are listed below. The efforts of these staff members have made issuance of this report possible.

Materials and Chemical Eneineerine Branch. Division of Encineerinc

Jack R. Strosnider Branch Chief 1 Robert A. Hermann Senior level Materials Engineer David Terao Chief, Chemical Engineering and Metallurgy Section Edwin M. Hackett Senior Materials Engineer, Materials Integrity Section James Medoff Materials Engineer, Materials Integrity Section William H. Koo Senior Materials Engineer, Chemical Engineering and Metallurgy Section Phillip J. Rush Materials Engineer, Inservice inspection Section Bamidele E. Akins Engineering Aide - Chemical Engineering COOP Student Mechanical Encineerine Branch. Division of Enrineerine Richard H. Wessman Branch Chief Kamal A. Manoly Section Chief, Component Integrity Section Jai R. Rajan Mechanical Engineer, Component Integrity Section C. Gary Hammer Mechanical Engineer, Component Integrity Section Mark Hartzman Senior Mechanical Engineer, Component Integrity Section 4

' Pat Patnaik Mechanical Engineer, Component Integrity Section 4

Reactor Systems Branch. Division of Systems and Safety Aswssment 4

Robert C. Jones Branch Chief Laurence E. Phillips Section Chief, BWR Reactor Section Kerri A. Kavanagh Reactor Systems Engineer, BWR Reactor Section Amy E. Cubbage Reactor Systems Engineer, Special Projects and Advanced Reactor Systems Section Probabilistic Safety Assessment Branch. Division of Systems and Safety Assessment Mark P. Rubin Section Chief, Program Integration and Application Section Proiect Directorate I-1. Division of Reactor Proiects 1/11 C. E. Carpenter Project Manager - Project Directorate 1-1 and lead Project Manager - BWR Internals Cracking Issue Don S. Brinkman Project Manager, Project Directorate 1-1 Consultants Carl J. Czajkowski Brookhaven National Laboratory E-1 NUREG-1544 4

N:grnRM 335 U S NUCLEAR REGUU. TORY COMMi$$10N 1QPORT E,R ,

EDE BIBLIO2RAPHIC DATA SHEET

'3"""""'"""'*"'*"*"'">

N U R E G-1544 l 2 TITLE AND SUDTITLE l Status Report: Intergranular Stress Corrosion Cracking of. BW R Core Shrouds and Other Internel Components 3 DATE REFORT PUBLISHED MONTH YkAH March 1996 4 FIN OR GRANT NUMBER l b AUTHOR ($) 6 TYPE OF REPORT 4

N R C Staff Report Technical i 1 PE RIOD COV E R E D sinclusow Onnent Septe m ber 30,1995 8 PERF MING ORGANiZ A T 10N - N AME AND AUDR E SS til Nnc pnwe owwon, Ortoco w neuen u 5 Nuen"at Neouaerary commossen. arus m.uione as,oren. osc ontractor, smwe l

Division of Engineenng '

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Com mission W ashington, D.C. 20555-0001 9 $PO R G G ANIZAT ION - N AME AND ADDR ESS fir N#c. trae ' Lee se eanw", .Icoarrator. paw ac Nac owsma. onw or s.,,u, u s Nocana n.puserary commes.es.

Same as above i

10 SUPPLEMENT ARY NOTES

0. Medoff, NRC Technical Monitor 11 ABSTRACT (200 aerds or nrass On July 25, 1994, the U.S. Nuclear Regulatory Com mission (N R C) 1ssued Gensnc Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water 4

reactors (B W Rs).

This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSC C) as it rehtes to the design and function of BW R core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the U.S. and abroad, followed by an indepth sum mary of the industry actions to address tha issue of IGSC C in B W R core shrouds and other internal components.

This report sum marizes the staff's basis for issuing G L 94-03, as well as the staff's assessment of plant-specific responses to G L 94-03. The staff is continually evaluating the hcensee inspection programs and the results from examinations of B W R core shrouds and oth:r internal components. This report is representative of submittals to and evaluations by ths staff as of September 30, 1995. An update of this report will be issued at a later date.

u u v wO R owoe SC R w i OR S no, w . ,.,,., ,a, w. .,m, , -~ ~ ,- ,- , . n .v.,mun uauMm Unlimited

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Boiling Water Reactor, Core Shroud, BW R Internal Components Int:rgranular Stress Corrosion Cracking, IGSC C, Genenc Letter 94-03 Uncl a ssified I v .. n a,1 Uncl a ssified Ib NUMBER OF PAGES 16 PRICE NRC PORM 3.5 Q HI

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NUCLEAR REGULATORY COMMISSION POSTAGE AND FEES PAfD ,

USNRC WASHINGTON, DC 20555-0001 PERMIT NO. G-67 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 4

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