RC-95-0173, Provides Plant Response to GL 95-03 Re Circumferential Cracking of SG Tubes

Provides Plant Response to GL 95-03 Re Circumferential Cracking of SG Tubes

ML20086B666
Person / Time
Site:	Summer
Issue date:	06/27/1995
From:	Gabe Taylor SOUTH CAROLINA ELECTRIC & GAS CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-95-03, GL-95-3, RC-95-0173, RC-95-173, NUDOCS 9507060033
Download: ML20086B666 (8)

Text

South Cir: lina Electric & GIs Ccmp:ny G:ry J.Tryl:r

. P.O. Box 88 Vice President Jenkinsville. SC 29065 Nuclear Operations (803) 345-4344 ~

SCE8G June 27,1995 essenc=m RC-95-0173 ,

Document Control Desk U. S. Nuclear Regulatory Commission

• Washington, DC 20555 Gentlemen: l

Subject:

VIRGIL C. SUMMER NUCLEAR STATION  !

DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 "CIRCUMFERENTIAL CRACKING OF STEAM GENERATOR TUBES i GENERIC LETTER 95-03"  !

I V. C. Summer Nuclear Station (VCSNS) understands the importance of performing i comprehensive examinations of steam generator tubes using techniques and i eq uipment capable of reliably detecting degradation to which the steam generator i tu bes may be susceptible. Recent operat,ng i experience with respect to the detection and sizing of circumferentialindications was evaluated at VCSNS. The replacement l of steam generators in December 1994 provides a new baseline for any past cracking .

or indications of steam generator tubes. The evaluation on the replacement steam  !

generators included a baseline consisting of full length bobbin coil for 100% of each  :

tube and 100% hot and cold leg transition zone measurements utilizin MRPC. We current have a total of 3 steam generator tubes plugged, all of hich were plugg preservice. l The replacement steam generators are tubed with Alloy 690 TT material which has )

been shown to be immune to PSWCC and highly resistant to other degradation 1 mechanisms. SCE&G will continue to utilize " state-of-the-art" NDE techniques to assess the condition of the replacement steam generators as was the practice with the original components. The eddy current inspection plan to be employed at our next refueling outage will meet or exceed those described in technical specifications l section 4.4.5.3.a.

VCSNS has developed plans for future steam c enerator tube inspections as they pertain to the detection of circumferential crac<ing. The scope (including sample expansion criteria) methods, equipment, and critena (including personnel training and qualification) are described in Attachment 1.

Should you have any questions regarding this submittal, please contact Mr. Michael Zaccone at (803) 345-4328.

Very truly yours, b  %

Garh ah!

MJZ/GJT/nkk Attachment

$\  :

03C050 t  !

ik 9507060033 950627 PDst ADOCK 0500o395 3-p PDR

' DocumGnt Control Desk

. LTR 950003 RC-95-0173 Ppge 2 of 2 4

c: J. L Skolds O. W. Dixon R. R. Mahan R. J. White General Managers S. Dembek NRC Resident inspector J. B. Knotts Jr.

J. I. Byrd R. E. Kelso J. S. Frick NEl RTS (LTR 950003)

File (815.14)

Central File System NUCLEAR EXCELLENCE- A SUMMER TRADITIONI

Att: chm:nt i

. LTR 950003 )

RC-95-0173 l Page 1 of 6 I

. 1 BACKGROUND l On April 28,1995, the U. 5. Nuclear Regulatory Commission (NRC) issued Generic Letter 95-03: "Circumferential Cracking of Steam Generator Tubes". Per GL 95-03, all addressees are requested tc,:

(1) Evaluate recent o aerating experience with respect to the detection and sizing of circumferentia indications to determine the applicability to their plant.

(2) On the basis of the evaluation in item (1) above, past inspection scope and results, susceptibility to circumferential cracking, threshold of detection, ex pected or inferred crack growth rates, and other relevant factors, develop a safety assessmentjustifying continued operation until the next scheduled steam generator tube inspections are performed.

(3) Develop plans for the next steam generator tube inspections as they pertain to the detection of circumferential cracking. The inspection plans should address, but not be limited to, scope,(including sample expansion criteria, if applicable), methods, equipment, and criteria (including personnel training and qualification).

This submittalincludes:

(1 ) A safety assessment justifying continued operation that is based on the evaluations performed in accordance with Requested Actions (1) and (2) above.

(2) A summary of the ins 3ection plans developed in accordance with Requested Actions (3) above anc a schedule for the next planned inspection.

Westinghouse Owner's Group generic response addressing' GL 95-03 requested actions entitled " Operational Data and Safety Assessment ,was used in the development of the V. C. Summer res sonse. Section 1.0 is an introduction section.

Section 2.0 is a listing of plants accoreing to tubesheet region expansion technique and tube material and history of circumferential cracking. Section 3.0 forms the basis for the safety assessment. Section 4.0 provides a defense in depth assessment.

1.0 INTRODUCTION

Recent nondestructive examination of the steam generator tubing at the Maine Yankee Nuclear Plant has identified a large number of circumferentialindications at the top of the tubesheet region. These most recent inspection findings, coupled with previously documented inspection results regarding circumferential cracking have lead to the issuance of NRC Generic Letter 95-03, "Circumferential Cracking of  ;

Steam Generator Tubes" on April 28,1995. The information detailed herein, will )

address the requested actions of the Generic Letter 95-03 as they pertain to l Westinghouse designed and manufactured steam generators in general and specifically to V. C. Summer. l The most recent inspection findings concerning steam generator tube expansion regions (Maine Yankee and Arkansas Nuclear One Unit 2) appear to have impacted I those steam generators utilizing the Combustion- Engineering (C-E) EXPLANSION i

Attrchm:nti

. LTR 950003 RC-95-0173 P, age 2 of 6 process more than others. While there are similarities between the C-E EXPLANSION  :'

process and the Westinghouse WEXTEX process, VCSNS utilizes the hydraulic expansion process, and Alloy 690 Thermally Treated tubes, which are immune to circumferential cracking experienced in the C-E units. Furthermore, the reported i sludge pile height at Maine Yankee (up to 18 inches) may have influenced indication detectability. Such sludge pile thicknesses are not representative of currently operating Westinghouse units.  ;

1.1 Historical Circumferential Degradation Locations Available historical information shows that for some Westinghouse plants, circumferential cracking has been detected in the tubesheet region tube expansion transitions from expanded to unexpanded tube, at the Row I and 2 U-bend tangent points, and at one plant (two twin units), at dented tube support plate intersections. ,

The main focus of this response will be to address tubesheet region expansion transition cracking, since this was the primary reason for the issuance of the generic letter. Other circumferential crack initiation sites will be addressed in the following section due to their limited numbers of field indications detected and limited number of tubes which can be affected (specifically small radius U-bends and dented l tube support plate (TSP) intersections).

1.2 Circumferential Degradation Evaluation of Small Radius U-bends and TSPs The incidence of circumferential indications at the Row I and 2 U-bend tangent i points has not been significant, both in numbers of indications at indicated RPC i angles. Some plants have administratively decided to preventively plug the Row 1, and in some plants Rows I and 2 tubes. Additional plants have applied U-bend heat ,

treatment in this region and have effectively recovered tubes previously preventively i plugged. Alloy 690 TT has not been shown to be susceptible to U-bend PWSCC.

A leakage event occurred in 1987 which resulted in a steam generator tube rupture due to high cycle fatigue at a dented top tube support plate. Pursuant to NRC Bulletin 88-02, all domestic Westinghouse steam generators with carbon steel tube

! sup port plates have been analyzed for the potential to experience high cycle fatigue -

1 at this location using a methodology accepted by the NRC. In cases where the analysis indicated that fatigue usage could exceed 1.0, the tube was either plugged and stabilized or lugged using a leak limiting sentinel plug. Three conditions must l

. be present for hi cycle fatigue at the top tube support plate; denting, lack of AVB l I support, and loc elevated steam velocities due to nonuniform AVB msertion depths. Steam generators with stainless steel support plates and broached

, (quatrefoil and trifoil) tube holes are not expected to experience this phenomenon.

An apparent ins section transient event occurred first in 1992 and involved the apparent identification of circumferentially oriented indications at the top of tubesheet region in a plant with partial depth roll ex pansion. Dec radation was detected using the bobbin probe, and denting was also associatec with many of the indications. Many of the indications had large voltage distorted bobbin indications, which is uncharacteristic of circumferential degradation. Several tubes were pulled i from the steam generator and destructively examined. The corrosion morphology was found to be closely spaced axial degradation and cellular degradation, as .

opposed to circumferential as suggested by RPC. In a cellular morphology closely 1

l l

t h- *

. . Attachment I-LTR 950003 RC-95-0173

[ Page 3 of 6 -

. spaced axial degradation and circumferentially oriented degradation can link and i

! form a tch like structure. The significant axial components of the cellular morph logy attributed to the distorted bobbin indication componentr TWoll specimens which were burst tested all produced axial burst o >enings, v  ? M sts  :

that the axial degradation dominated the morphology. The icensee pedoi, o  ;

100% bobbin and 100% RPC inspection of the top of tubesheet region of all steam

.j generators in 1992 and also in 1994. -

i:' Circumferential cracking at dented TSP intersections has been detected at one plant  !

(two twin units). The steam generators experiencing this phenomenon have been i replaced. This enomenon has not been detected at other units. This plant also  !

' operated at hi her temperatures than most other units.  !

I In many Westinghouse plants, an augmented top of tubesheet region inspection l j program is conducted on a cycle to cycle basis. Many Westinghouse plants have had i

- all hot leg tubes inspected at the top of tubesheet region using the RPC probe and i

{. continue to do so on a cycle to cycle basis. Currently available probes, coupled with .

[

properly implemented reporting criteria and techniques have been demonstrated to be sufficient to identify circumferential indications in the tubesheet region.

j i Recognizing the potential susceptibility to cracking in the expansion transition .

regions, many Westinghouse units have implemented shotpeening or rotopeening  :

i- of the expansion transitions to enhance the resistance of this region to the tube j - bundl,e to primary water stress corrosion cracking (PWSCC). This remedial measure, ,

especially when implemented prior to commercial operation, can be effective in

e mitigating the effects of PWSCC. i L

i, Collectively, provide justi fication for the continued operation of V. C. Summer.the items discus

[

i-

! 2.0 OPERATING EXPERIENCE WITH CIRCUMFERENTIAL CRACKING FOR'THE U.S.

j POPULATION OF WESTINGHOUSE STEAM GENERATORS SIMILAR TO VCSNS

The V. C. Summer Nuclear Station uses Westinghouse Delta 75 steam generators.

i Other plants with the types of steam generators using a similar tube expansion L process are Cook Unit 2, Indian Point 3, and North Anna Units 1 & 2. The steam .

i. generators at VCSNS use Alloy 690 TT tubing. The nominal tube OD is 11/16 inch OD j x 0.040 inch nominal wall thickness. There has been no evidence of circumferential

^

crackincl in any location of Alloy 690 Thermally Treated (TT) tubing utilizing

[ hydrauIcally expanded joints.

4 3.0 SAFETY ASSESSMENT SUPPORT FOR ALLOY 600 TT AND ALLOY 690 TT TUBE i MATERIAL PLANTS i  :

[ Thermally treated Alloy 600 tubing represents an intermediate step in the evolution

of progressively optimized corrosion resistant tubing materials. EPRI report NP-3501,

" Optimization of Metallurgical Variables to improve Corrosion Resistance on inconel

- Alloy 600" shows the distinct advantage of Alloy 600 TT over Alloy 600 Mill

. l I. Annealed. Data contained in this report shows minimal SCC in 600 TT c-rings at 600* i j F in caustic solutions (10% NaOH).' Crack depths were generally 2.5 to 4.5 times less  ;

e than Alloy 600 MA at 600* F. Primary water SCC initiation times were also found to L 4 1

4

[

v e>=+ +w.,-. -=m ee.-=--s rrv e -r e-- e +-a-,m-<s-o.-e,v-.c---irea.m-eim .--e,4w-=we no +- m e-em - e w i v e -e se we e -+ -we r- e n wr w ee e~ e- e+si.-t+-e+--n-n---r-wetee,-t<es-e -w w w-

I -

. - AttachmsntI i .- LTR 950003 e '

RC-95-0173 Page 4 of 6 1

. be greater for 600 TT versus 600 MA. Also, no SCC was detected in Alloy 600 TT small radius U-bends tested at 680 F. This report also showed a dependence upon residual stress level and crack growth rate and initiation times. Westinghouse data has j , shown that the stress levels in hydraulically expa,nded tubing are less than the associated levels in either explos,vely i or mechanically expanded tubes. Also, some 4 plants utilizing Alloy 600 TT (with hydraulic ex sansion) operate at significantly lower I temperatures, about 15 to 20 degrees lower, t 1an the plants with Alloy 600 MA i hydraulically expanded tubing. Since corrosion rate is temperature dependent, a t lesser potential for rapid corrosion would be expected.

} Alloy 690 TT tubing represents a further advancement in the evolution of 4 progressively optimized corrosion resistant tubing materials above Alloy 600

thermally treated. Testing programs have indicated that Alloy 690 TTtube material provides for significant PWSCC resistance and an increase in ODSCC resistance, compared to Alloy 600 TT. Alloy 690 TT is generally accepted as the steam generator i ~ tube material of choice. Alloy 690 TT tubinc bas been in service at Cook Unit 2 since i- 1989 with no reported instances of localizec tube wall degradation. Westinghouse

{ Alloy 690 TT sleeves have been in service since 1983. There has been no reported

[ degradation in these sleeves.

4

! 3.1 ~ Pulled Tube Examination Results and Tube Integrity Assessments Performed by

Westinghouse

in 1990 two tubes, R10 C53 and R25 C57 were pulled from the Surry Unit I

replacement steam generators (Alloy 690 TT tubing). Field NDE suggested the i presence of circumferentially oriented d concluded that the poorly defined RPC s,egradation. Upon further review it wasig

" ding" or mechanical deformation. Upon destructive examination, no corrosion, either ID or OD initiated was detected. The source of the NDE indications was

. determined to be attributed to probe liftoff in the expansion transition and i mechanical conditions in the tube resultant from the tube installation process. The

maximum dianieter of R25 C57 occurred approximately 0.6 inch above the top of the tubesheet. A 70* " groove", mechanical in nature was found on the tube OD and i attributed to the interaction of the tube with the edc e of the tubesheet during the

~

expansion process. It has been concluded that this tu ae was overexpanded above the top of the tubesheet. The hydraulic expansion process used was designed to locate the transition slightly below the top of the tubesheet.

l

'~

Due to the lack of circumferential indications in this type of tubesheet region expansion technique, Westinghouse has not performed any plant specific tube integrity evaluations for full depth hydraulically expanded Alloy 690 TT tubing.

3.2 Inspection Methodologies and Adequateness V. C. Summer Nuclear Station (VCSNS) understands the importance of performing comprehensive examinations of steam generator tubes using techniques and equipment capable of reliably detecting degradation to which the steam generator tubes may be susceptible. Recent operating experience with respect to the detection and sizing of circumferentialindications was evaluated at VCSNS. The replacement of steam generators in December 1994 provides a new baseline for any past cracking

~ - -.

"i Attrchm::nt i .

U - LTR 950003 l l

RC-95-0173 Page 5 of 6 3:

or indications of steam generator tubes. The evaluation on the replacement steam e'

generators included a baseline consisting of full lencith bobbin coil for 100% of all l tubes in each steam generator and 100% hot and co d leg transition zone i measurements utilizmg MRPC. We currently have a total of 3 steam generator tubes  !

!. plugged, all of which were plugged preservice. j i The replacement steam generators are tubed with Alloy 690 TT material which has i l been shown to be immune to PSWCC and highly resistant to other degradation 1

mechanisms. - SCE&G will continue to utilize " state-of-the-art" NDE techniques to

i assess the rondition of the replacement steam generators as was the practice with -

i the original components. i i -

~

l

} 3.3 Inspection Personnel Qualification & Training j l Steam Generator tube examinations at VCSNS are performed by a qualified vendor. l

Inspection & Analysis personnel are trained and certified m accordance with  !

t American Society of Non-destructive Testing, Recommended Practice SNT-TC-1A, '

i 1984 edition. Additionally, analysis personnel are required to attend and i satisfactorily complete a site specific training program prior to start of tube i

, inspection activities. This training program incorporates the classroom portion of l

! this requirement described in EPRI document NP-6801 Revision 3 titled " Steam l Generator tube Examination Guidelines" appendix 'G', as applicable. This training ,

program provides each analyst with site operating history and analysis '

i metlodologies used at VCSNS. 4 i

l 3.4 Scope of Examination for Next Refueling Outage i The baseline examination of the Delta 75 stearp generators was serformed using 1

state-of-the-art" equipment and current accepted methodo ogies for data

~

acquisition and analysis. The baseline examination was performed as described in VCSNS technical specifications section 4.4.5.4.a.9. ,

The scope of examination and augmented inspection plans for the next refueling ,

outage (Spring '96) will meet or exceed those described in technical specifications section 4.4.5.3.a.

3.5 Individual Plant Tube Integrity Assessments The past two inspection programs at V. C. Summer were performed on the originally installed steam generators. These inspections are not considered relevant to the generic letter response due to the current configuration at V. C. Summer. The evaluation on the current, replacement steam generators included a baseline consisting of full length bobbin coil for 100% of each tube and 100% hot and cold leg transition zone measurements utilizing MRPC. These inspections are consistent with the EPRI and industry guidelines regarding reporting criteria guidelines and initial sample inspection size such that any structurally significant circumferential indications would have been identified. There are no domestic operating experiences of circumferential indications'in full depth hydraulically expanded plants with Alloy 600 TT or Alloy 690 TT tubing. No indications are expected at V. C.

1 f . '-

~ ,.

Attichmcnt l' LTR 950003

! RC-95-0173 -  !

P, age 6 of 6 .)

o .

Summer wh'ich would challenge tube integrity at the end of the current operating cycle. Similarly, tube structural integrity is expected to be maintained during all

future operatmg cycles considering the design of these expansions, and also i assuming that plant operating parameters will not be significantly altered from j current conditions.

]

4.0 DEFENSE IN DEPTH ASSESSMENT POINTS I

4.1 Alloy 600 TT and Alloy 690 TT Tube Material

{

Plants with Alloy 600 TT tubing utilize full depth hydraulic expansion. The a_ pparent i 3

lack of susceptibility to rapid degradation of hydraulically expanded Alloy 600 TT  !

{ tubing is seen by the operating experience of plants using Model F steam

- . generators. Plants with Alloy 600 TT tubing have been operating since 1980 with no <

i- reports of corrosion degradation. Plants with Alloy 690 TT tubing also utilize full .

[ ' depth hydraulically expanded tubing. Alloy 690 TT tubing has been shown by

! extensive testing programs to re 3 resent the " state-of-the-art" in corrosion resistant

. steam generator tubing materia . -There is no evidence which suggests that rapid i i corrosion degradation of Alloy 600 TT tubing or Alloy 690 TT tubing would be i experienced, either up to the end of the current operating cycles for these units, or

during any cycle in the near future. ,

i.  !'

1.

j 4.2 EOPs 1

l The emer cy operating procedures are specifically designed to respond to single and mult tube rupture scenarios. The NRC has performed additional analysis efforts (o tlined in Draft NUREG-1477 and NUREG-0844) which indicate that the refueling water storage tank (RWST) would not become depleted during response to multiple tube rupture events. ,

l l

L i

l c _____ - .._ .. -- . . . - - - . . , - . - - - - - . - . - - . . _ _ - . , , - , . . . . - - - - . . . . . - . ~ . _ , . . , . . . . . . - , , , - . - , , ,