ML20217N302

Insp Repts 50-313/98-12 & 50-368/98-12 on 980302-11. Violations Noted.Major Areas Inspected:Maint & Engineering

ML20217N302
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/02/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20217N251	List: ML20217N248 ML20217N271 ML20217N302
References
50-313-98-12, 50-368-98-12, NUDOCS 9804090068
Download: ML20217N302 (19)

See also: IR 05000313/1998012

Text

m

,

.

.

ENCLOSURE 2

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket Nos.:

'50-313;50-368

License Nos.:

DPR-51; NPF-6

Report No.:

50-313/98-12;50-368/98-12

Licensee:

Entergy Operations, Inc.

Facility:

Arkansas Nuclear One, Units 1 and 2

Location:

Junction of Hwy. 64W and Hwy. 333 South

- Russellville, Arkansas

Dates:

March 2-11,1998

Inspectors:

1. Bames, Technical Assistant

C, A. Clark, Reactor inspector, Maintenance Branch

> Accompanied By:

Dr. C. V. Dodd, NRC Consultant

I

Approved By:

Dr Dale A. Powers, Chief, Maintenance Branch

Division of Reactor Safety

i

ATTACHMENT:

Supplementalinformation

q

i

l

i

i

i

. , .

1

j

1

i

9804090068.900402

1

PDR

ADOCK 05000313

"

l

G-

PDR

j

p

.

i'

4

2-

EXECUTIVE SUMMARY

Arkansas Nuclear One, Units 1 and 2

NRC Inspection Report 50-313/98-12; 50 368/98-12

Maintenance

The observed eddy current data examinations were performed in a thorough manner..

Examiners were knowledgeable of their assigned tasks (Section M1.1).

The licensee's third 10-year inservice inspection program plan inspection interval started

June 1,1997, and the licenses was still developing the program plan. The licensee had

submitted an initial portion of the third 10-year inservice inspection program plan to the

Office of Nuclear Reactor Regulation for review (Section M3.1).

The reviewed inservice inspection examination records were documented

appropriately (Section M3.2).

A noncited violation was identified for the failure to comply with Section XI ASME Code

Examination requirements to property distribute examinations in the Unit 1 second 10-

year inservice inspection interval (Section M3.3).

The examiners scheduled to performed nondestructive inservice inspection examinations

during Refueling Outage 1R14, were certified appropriately (Section M5.1).

' ED91DMIDS

The licensee was continuing to satisfactorily implement the eddy current examination

program enhancements that were !nitially introduced during Refueling Outage 2R12

(Section E8.1).

A violation of Criterion IX of Appendix B to 10 CFR Part 50 was identified pertaining to

the failure to assure use of a phase rotation setting, for analysis of plus point probe eddy

current data, that was consistent with the setting that was used in the Appendix H (of the

Electric Power Research Institute "PWR Steam Generator Examination Guidelines")

. qualification of the procedure (Section E8.1).

'-

A violatic q c' Criterion V of Appendix B to 10 CFR Part 50 was identified pertaining to a

l

change mid's to the calibration requirements of Engineering Standard HES-72, "ANO

Eddy Curn;nt Data Acquisition," Revision 0, without use of either a standard change

,

notice or re. vision of the engineering standard (Section E8.1).

1

,-),

%

t

---

i

.

-3-

Reocit Details

Summary of Plant Status

Unit 1 was operating at 100 percent power and Unit 2 was in a planned mid-cycle outage 2P98

during the onsite portion of the inspection.

II. MaintenRDER

M1

Conduct of Maintenance

M1.1 Inservice insoectirn (73753)

a.

Insoection Scoce

The inspectors observed Unit 2 eddy current data acquisition for various nondestructive

steam generator tube examinations performed on March 3,1998, in the hot and cold legs

of Steam Generators A and B. The observations of 6 examiners performance, were

performed at various times during the day shift, for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of observation

time. This included observations of calibration of acquisition system, verification of the

robotics manipulator arm position, data acquisition for both bobbin coil and motorized

rotating pancake coil examinations, and contractor initial review of acquired data during

performance of examinations.

b.

Observations and Findinos

Engineering Standard HES-72, "ANO Eddy Current Data Acquisition Guideline,"

Revision 0, specified the licensee process for multi-frequency eddy current testing of the

tubing, sleeves, and plugs of ANO-1 and ANO-2 steam generators. The inspectors

reviewed procedure HES-72. The inspectors discussed procedure HES-72 and

observed data acquisition examinations with both licensee and contractor personnel.

The inspectors found the observed eddy current examinations were performed in

accordance with the applicable procedures and ASME Code requirements.

Discussions with the examiners performing the eddy current examinations indicated that

thof were experienced and knowledgeable nondestructive examination personnel. The

observed examiners were cognizant of the procedural and documentation requirements,

and understood the examination processes and techniques.

c.

Conclusions

The observed eddy current data examinations were performed in a thorough manner.

Examiners were knowledgeable of their assigned tasks.

'

.

_c

-4-

>

J

M3:

Maintenance Procedures and Documentation

M3.1l Inservice Inspection Program Plan and Schedule

s.

Inspechon Scope (73753)

The inspectors reviewed Revision 0 of the licensee's Unit 1 third 10-year inspection

interval (June 1,1997, through May 31,2007) inservice inspection program plan. The -

inspectors also reviewed Revision 4 of the schedule of proposed inservice inspection

examinations scheduled to be performed during Unit 1 Refueling Outage 1R14, which

was scheduled to start March 28,1998. These documents were reviewed to determine

if changes to the Unit 1 inservice inspection program plan concerning component.

selection, etc., had been properly documented and approved consistent with the

'

requirements of Section XI of the ASME Code,1992 Edition, without Addenda. The

inspectors discussed the Unit 1 third 10-year inservice inspection program plan and the

Refueling Outage 1R14 inservice inspection f.chedule with personnel in engineering

. programs and nondestructive examination,

b.

Observations and Findings

The inspectors determined that the Office of Nuclear Reactor Regulation was currently

reviewing the licensee's Unit 1 third 10-year inservice inspection program plan, which

was submitted June 25,1997, to determine the plan compliance with 10 CFR 50.55a(g).

The inspectors noted that on December 2.1997, the Office of Nuclear Reactor

Regulation submitted a request for additional Unit 1 third 10-year inservice inspection

program plan information such as: (1) system boundary diagrams, (2) isometric and/or

- component drawings, (3) summary of examination (itemized listing of the components)

1

. scheduled to be performed during each period of the third 10-year interval, and (4) other

identified matters pertaining to the program plan. In discussions with representatives of

the licensee's engineering programs, the inspectors noted that the licensee was in the

process of developing the information requested by the Office of Nuclear Reactor

Regulation, and had not submitted the requested information as of March 6,1998.

While Revision 0 of the Unit 1 third 10-year intervalinservice inspection program plan

described the ASME Code Class 1,2, and 3 components / examination categories subject

<

rc

to surface, volumetric, and visual examinations, and noted various relief requests for

,.

identified areas, the plan was still in the developmental process. The licensee was

aware that it had approximately three and half years in the first inspection period of the

i

'

- third 10-year inservice inspection interval, to resolve any inservice inspection program

plan questions.'

L

i c."

Conclusions

'The licensee #s third 10-year inservice inspection program plan inspection interval started

m

June 1,'1997, and the Ocensee was still developing the program plan. The licensee had

m-

,

.

,

g.

A

% /

y.

3

e

,

l

.,

.-

-5-

I

l

submitted an initial portion of the third 10-year inservice inspection program plan to the

Office of Nuclear Reactor Regulation for review.

{

'

M3.2 Inservice insoection Proaram Examination Records

a.

Insoection Scooe (73753)

' The inspectors reviewed completed examination records for 46 nondestructive inservice

inspection examinations (listed in the attachment to this inspection report) performed

j

during Unit 2 Refueling Outage 2R12. The examination records were reviewed to verify

I

that examination activities and results were documented in accordance with the

licensee's program, procedure, and ASME Code requirements.

b.

Observations and Findinas

The inspectors determined that the reviewed examination records had been prepared in

accordance with licensee program, procedure, and ASME Code requirements.

c.

Conclusions

The reviewed inservice inspection examination records were documented appropriately.

1

M3.3 Inservice insoection Proaram Plan Reauirements

a.

Insoection Scoce (73753)

The inspectors reviewed licensee activities for evaluating the Unit i second 10-year

inservice inspection examination data, to ensure examinations were performed in

accordance with the requirements of Section XI of the ASME Code and

- 10 CFR 50.55(a),

b.

Observations and Findinas

The inspectors noted that during review of nondestructive examinations performed

for the Unit 1 second 10-year inservice inspection interval, the licensee discovered

that inspection period distribution requirements had not been met for certain

- Section XI ASME Code Examination Categories. Additionally, this review rev6aled

that, as a consequence of improperly distributing examinations, only 35 of the

39 required Section XI ASME Code Examination Category C-C integral attachment

= weld examinations had been performed. The licensee issued Condition

. Report CR-ANO-1-1998-0127, to document this failure to inspect in accordance with

.

the Section XI ASME Code requirements. A subsequent operability assessment found

- allimpacted components remained operable. This failure constitutes a violation of minor

significance and is being treated as a noncited violation, consistent with Section IV of

the NRC Enforcement Policy (50-313/9812-01).

.}

.

.

t

.

, r

.

.7

_ -6 '

I

The inspectors reviewed Condition Report CR-ANO-1-1998-0127 and discussed it with.

-l

.

-

'

-- engineering programs personnel. The inspectors noted that the licensee implemented

corrective actions to submit a relief request to the NRC, inspect the four missed -

1

Category C-C welds during the next refueling outage, and to ensure that all required

i

examinations were scheduled and distributed appropriately for the third 10-year inservice

-inspection interval. The inspectors verified the four missed Category C-C welds were .

'

identified in Revision 4 of Refueling Outage 1R14 schedule for inservice inspection

outage examinations.'

c.

Conclusions

A noncited v'iolation was identified for the Unit 1 second 10-year inservice inspection

interval for failure to comply with certain Section XI ASME Code Examination

requirements.

'

M6

Maintenance Staff Training and Qualification

MS.1 - Personnel Qualification and Certification

-

a. s . Insoection Scone'(73753)

The inspectors reviewed the qualifications and certifications of the inspection personnel

involved with the inservice inspection program, to verify that the certification process met

-

. the requirements of American Society for Nondestructive Testing's " Recommended

Practice SNT-TC-1 A," 1984 Edition. The inspectors reviewed the available qualification

files for one nondestructive examination Level ll examiner and two Level lil examiners

scheduled as of March 5,1998, to perform examination activities during the Unit 1

Refueling Outage 1R14, scheduled to start March 28,1998.

b;

Observations and Findinas

The licensee representative informed the inspectors that the performance of inservice

' inspection examinations were contracted out, but were performed by nondestructive

examination examiners whose certifications were reviewed and approved by the

licensee. The inspectors observed that the site-Level lli examiner exhibited a high

degree of competency and was fully cognizant of ASME code requirements and

inservice inspection program commitments. The inspectors noted that the personnel

qualification and certification files reviewed contained the appropriate examinations and

certifications for the designated nondestructive examination methods. The records

indicated that personnel had been certified in accordance with the Recommended

Practice SNT-TC-1 A,1984 Edition. As required by the ASME code, all of the individuals

"

had maintained current documontation regarding near-distance acuity and color vision

examinations.-

.

l k1

'l k r

'

f

i

'

g[i'

M

, ,

4

1

.

,

-7-

c.

CDnclusions

The examiners scheduled to perform examinations during Refueling Outage 1R14, were

]

certified appropriately.

M8

Miscellaneous Maintenance issues (92902)

M8.1 (Closed) Violation 50-313:-368/9713-01: Failure to include certain safety-related valves

in the inservice test program, and failure to appropriately test certain valves in their safety

function position.

l

Valves BW 4A and BW-4B (Unit 1 borated water storage tank outlet check valves)

were included in the inservice test program; however, they were identified as having

an open safety function only, and the closure function was not being tested. Check

Valves 2BS-1 A and 2BS-1B (Unit 2 Refueling Water Tank Outlet Check Valves) were

j

'

similarly identified. Further review by licensee personnel revealed four additional check

valves that were identified as not having a closed safety function, yet were considered

part of a dual isolation configuration (CA-61, CA-62 - sodium hydroxide tank outlet check

valves, and BW-2, BW-3 - high pressure injection pump suction check valves). Licensee

personnel conducted an operability assessment on these valves and determined them to

be operable based on recent surveillance test information and periodic maintenance.

The inspectors reviewed and agreed with the assessment.

The licensee's failure to test or exercise the above eight valves to verify their ability to

fulfill all safety functions, was identified as a violation.

Further licensee review identified an additional seven Unit 2 ASME Code, safety-related,

normally closed valves that have an open safety function, but were not in the inservice

test program. The valves were identified as 2FP-31,2FP-46,2SW-138,2SW-56,

2SW-57, 2SW-62, and 2SW-67, all Category B valves in the service water piping, which

provides makeup water to the spent fuel pool. Licensee personnel performed an

operability assessment on these valves and determined that they were operable based

on recently performed surveillance tests on other equipment, which required opening of

the seven valves.

The licensee's failure to include ASME Code, safety-related valves in the inservice test

program was an additional example of a violation of 10 CFR 50.55a(f)(4), and Section XI

of the ASME Boiler and Pressure Vessel Code.

-The licensee's staff implemented actions to: (1) include all ASME Code, safety-related

valves in the inservice test programs for both units, and (2) test, exercise, or examine all

identified ASME Code valves to verify their ability to fulfill all safety functions. The

inspectors reviewed licensee's records, procedures, and other documents and fou,1d that

the licensee had implemented appropriate corrective action for this violation.

-8-

M8.2 (Closed) Deviation 50-313/9713-02: Failure to meet commitments regarding inservice

inspection frequency with no subsequent notification made to NRC.

On April 21,1997, Duke Power Company, the licensee for Oconee 2, identified leakage

in a high pressure injection nozzle. The leakage appeared to be the result of fatigue with

flow-induced vibration as a likely contributor. The unisolable pressure boundary leak

was a precursor to a small break loss-of-coolant-accident. This failure mechanism was

identified in 1982 at Crystal River, Unit 3, which experienced an unexplained loss-of-

coolant on January 24,1982. Subsequently, it was revealed that the high pressure

injection / makeup nozzles were cracked. The NRC issued Information Notice 82-05. The

licensees who were users of Babcock & Wilcox nuclear steam supply systems formed a

B&W Owners' Group Safe-End Task Force that established a root cause and made

recommendations to address the problem. This problem became Generic issue 69 and

the NRC issued Generic Letter 85-20, which endorsed the Owners' Group

recommendations.

The licensee committed to perform Recommendation 3, which was then added to the

inservice inspection program augmented testing requirements for the injection nozzles.

By letter dated April 22,1985, the licensee informed the NRC of its agreement to

implement the recommendations of the B&W Owners' Group. The licensee had, in fact,

already initiated implementation of the augmented testing during the Unit i forced outage

in 1982.

In 1989, the licensee identified that certain planned examinations had been missed. This

was documented in Condition Report 1-89-0508. The examinations were identified in the

j

inservice inspection program as augmented examinations, rather than commitments to

the NRC. As a result of the categorization of the examinations, subsequent work

schedule or ALARA demands led to the examinations being canceled.

The corrective actions identified in Condition Report 1-89-0508 resulted in the

performance of radiographic and ultrasonic examinations during Refueling Outage 1R9.

Licensee personnel also performed an ultrasonic examination of Nozzle D during

Refueling Outage 1R12 in February 1995. In addition, the licensee established plans to

'

perform radiographic and ultrasonic examinations of all nozzles during the spring of 1998

(Refueling Outage 1R14).

However, between conduct of the initial committed radiographic and ultrasonic

,

examinations during Refueling Outage 1R5 in February 1982 and Refueling Outage 1R9

in November 1990,12 of the 14 scheduled examinations were not performed. The

licensee failed to meet the commitments made to the NRC in the April 22,1985, letter

(1CAN048501) and the NRC was not informed of a change to the commitment. This

failure to implement a commitment was identified as a deviation.

i

m

j

.

.g.

The inspectors reviewed the licensee July 28,1997, response to this deviation and

discussed the corrective actions taken with the licensee's staff. The inspectors reviewed

print outs of applicable sections of the computer data base for the Unit 1 third 10-year

inservice inspection program plan, since the licensee had not issued a formal completed

copy of the plan as of March 6,1998. The inspectors verified that the licensee had

revised the Unit 1 third 10-year inservice inspection program computer data base to

include specific criteria for examination of the thermal sleeve to safe end area for gaps

on the high pressure injection / makeup nozzles. The inspectors noted that the

augmented examinations for the high pressure injection / makeup nozzles were scheduled

for Refueling Outage 1R14 and at a frequency of every five refueling outages thereafter.

The inspectors found that the licensee had implemented appropriate corrective action for

this deviation.

111. Enaineerina

E8

Miscellaneous Engineering lasues (92700 and 92903)

E8.1

(Ocen) Violation 50-313: 368/9714-01: (1) Unit 1 potential for steam generator tubes to

be left in service that exceeded the plugging limit of technical specifications, and

(2) Unit 2 tubes that had defects in excess of the plugging limit of the technical

specifications.

j

On April 8,1997, the licensee discovered (from metallographic examination of three tube

samples removed during Unit 1 Refueling Outage 1R13) that an eddy current technique

that had been employed during the refueling outage to size depth of intergranular attack

showed a nonconservative bias of up to 50 percent through-wallin the measurements,

thus, creating the possibility that tubes could have been left in service with flaws, which

exceeded the plugging or repair limit of the technical specifications. On April 11,1997,

the NRC staff made a determination to exercise discretion not to enforce compliance with

Technical Specification 4.18.5.b until the earlier of May 7,1997, or the date of issuance

of an amendment to Technical Specification 4.18.5.b. The amendment, which was

issued May 7,1997, authorized operation to the next refueling outage with tubes

containing intergranular corrosion indications in excess of the technical specifications

plugging limit.

l

The inspector verified that the licensee had taken administrative action to preclude future

use of the sizing technique for intergranular attack at Arkansas Nuclear One.

Specifically, Standard Change Notice 04 to Engineering Standard HES-27, 'ANO-1

Steam Generator ECT Data Analysis Guidelines," Revision 4, precluded use of the

suspect regression analysis technique after Refueling Outage 1R13. The licensee

additionally identified in its December 18,1997, response to the Notice of Violation that

the Unit 1 steam generator tubes, with indications of intergranular attack in the upper

__

)

,

,

,

.

,

\\

-10-

,

tube sheet area, would be dispositioned prior to startup from the next refueling outage.

1The licensee was informed during the exit meeting on March 6,1998, that the Unit 1 >

violation would remain open pending review of the disposition of.the intergranular attack -

,

indications. The licensee was requested to submit the disposition information upon it

"

becoming available during the upcoming Refueling Outage 1R14.

With respect to the Unit 2 part of this violation, the inspectors reviewed the' current eddy

current examination program requirements and observed Outage 2P98 eddy current :

. acquisition and ' analysis activities, in order to verify that the licensee was continuing to

.

'

implement the program enhancements that were initially introduced in Refueling Outage

2R12. The inspectors noted that the program requirements were comprehensive and

comparable to those used in Refueling Outage 2R12. The licensee was contiriuing to

use an "Entergy Review Group" for independent oversight of eddy current analysis ,

performance. The group charter included review of data discarded by the resolution

process, sampling of tubes with fno detectable' degradation" calls by primary and

secondary analysts, review of guideline change forms, and ensuring that all repairable

,

codes that had been dispositioned as non-repairable by resolution analysis had been

, ,

appropriately reviewed and documented. The inspectors additionally verified that

.

licensee personnel were directly involved in the training of analysts, with testing heavily

focused on known Unit 2 degradation modes. The licensee was noted to be continuing

to use the Zetec analyst performance tracking software for tracking, trending, and

feedback to analysts of analysis errors. Overall, the inspectors concluded that the

licensee was continuing to satisfactorily implement the program enhancements that were

introduced in Refueling Outage 2R12.

During the review of the Outage 2P98 eddy current examination program, the inspectors

noted that one of the licensee approved eddy current technique specification sheets (i.e.,

ETSS #4), which had been prepared for plus point probe examinations of dented

locations, contained a technical deficiency. Specif;cally, ETSS #4 instructed the analyst

-

to adjust phase rotation so that probe motion was horizontal. The inspectors considered

that this guidance was technically inappropriate for the plus point probe, due to its

insensitivity to probe motion resulting in too small a signal to allow this adjustment to be

accurately accomplished. The inspectors noted that improper setting of phase rotation

could also negatively impact the ability to detect small inside diameter flaw indications.

The inspectors requested to see the supporting Appendix H (of the Electric Power

Research Institute "PWR Steam Generator Examination Guidelines") qualification for

- ETSS #4 and were provided ETSS # 96402 odsec2. doc (from the Electric Power

Research institute Performance Demonstration Data Base)in response. The inspectors

'

ascertained from review of ETSS # 96402 odscc2. doc that ETSS #4 was 'not consistent

with its qualification (i.e., ETSS # 96402 odscc2. doc specified a phase rotation setting of

15* for 40 percent through-wall circumferential and axial inside diameter notches).

1

Licensee staff initiated Condition Report CR-ANO-2-1998-0090 in response to the

_

identified deficiency in ETSS #4 and informed the inspectors that the eddy current

L technique specification' sheet had not been used during Outage 2P98 for eddy current

, ,

!

'

n

-

.

,

et

,

a;

- . .

7

-11-

data acquisition and analysis. The inspectors were also informed that a similar condition

had previously existed with respect to ETSS #3. This condition was ide,1tified and

corrected prior to use of ETSS #3. Licensee staff informed the inspectors that they

.

' believed the error in ETSS #4 would also have been identified and corrected prior to its

use. Licensee staff were informed during a final telephonic exit meeting on March 11,

1998, that the failure to assure use of the Appendix H qualified phase rotation was a -

potential violation of Criterion IX of Appendix B to 10 CFR Part 50 (50-368/9812-02);

' As part of the assessment of Unit 2 eddy current' analysis performance, the inspectors

p

selected a limited sample of Unit 2 Outage 2P98 eddy current data for independent

- review by the NRC consultant. The sample scope consisted of one tube, which had -

been identified to contain an axial free span indication, and tubes that had been identified

to contain small-to-moderate amplitude axial flaw indications at eggerate locations. The

.

NRC consultant reviewed both the bobbin coil and 0.115-inch motorized rotating

. pancake coil data obtained in Outage 2P98 for the sample, and also the available

!

eddy current data for the sample tubes that was obtained during the prior Refueling

]

Outage 2R12. The NRC consultant had no significant disagreements with the " calls"

made during either Refueling Outage 2R12 or Outage 2P98. During the review, the NRC

consultant examined the calibration readings for ten calibration groups. The motorized-

rotating pancake coil readings for the 100 percent through-wall axial notch were

observed to be saturated (i.e., no further increase in digital output as analog input signal

was increased) for all groups. The corresponding calibratior.s from Refueling Outage

2R12 were checked and noted to be not saturated. The NRC consultant concluded that

the use of a saturated calibration signal would not affect the ability of the motorized

rotating pancake coil to detect defects, but would result in errors in measured voltages

and depths of flaws. The inspectors were informed by Framatome eddy current

personnel that the potential for a saturated signal from the 100 percent through-wall

notch was known prior to Outage 2P98, as a result of testing performed at another plant.

Preliminary eddy current examinations using the Arkansas Nuclear One, Unit 2,

1

calibration standards confirmed that a saturated signal was obtained when a 100 percent

through-wall axial notch was set at 20.0 volts. To avoid normalizing to a voltage value

_

- from a saturated signal, Framatome developed an attemate method. This method

required setting the signal amplitude from a 60 percent through-wall inside diameter

.. notch to 7.0 volts, when a saturated signal was obtained from the 100 percent through-

. wall axial notch. Licensee personnel informed the inspectors that the change in voltage

normalization practice had been discussed during training of the analysts, and added as

'

-' a footnote to the "ANO-2 Calibration Standard As-Built Dimensions" sheet that was -

provided to analysts for use in performing setup and calibration. The NRC consultant

noted, however, during review of Calibration Group 0059 that the voltage appeared to

i

have been normalized using a saturated signal from the 100 percent through-wall axial

l

notch; rather than the altemate 7.0 volt setting from a 60 percent through-wall inside ;

j

. diameter notch.

'\\

,i

1

t

4I

i j ,.

> >

i

d

<

i

,

i

n

s

.~

~

.s.

,

,

.-12-

The inspectors questioned licensee personnel regarding whether the use of the attemate

calibration method had been approved by the licensee in accordance with eddy current

examination program requirements. Specifically, paragraph 5.4.5 in Engineering -

Standard HES-72, "ANO Eddy Current Data Acquisition," Revision 0, requires that

changes made to the standard are to be completed using either a standard change

notice or by revision of the standard. The inspectors were subsequently informed that

- the change in calibration practice was not completed by either of these two methods.

Licensee staff initiated Condition Report CR-ANO-2-1998-0089 in response to the -

identified deficiency. Licensee staff were informed during a final telephonic exit meeting

on March 11,1998, that the failure to formally incorporate the change to Engineering

Standard HES-72, Revision 0, by use of either a standard change notice or revision

'

of the engineering standard, was a potential violation of Criterion V of Appendix B to

10 CFR Part 50 (50-368/9812-03).

E8.2 (Closed) Wlation 50-368/9714-03* Lack of prompt corrective action in 1995 prior to

. returning potentially defective sleeved tubes to service..

The inspectors examined the results of an NRC staff review of weld defects in

Combustion Engineering steam generator tube sleeves that was sent to the licensee by

letter dated March 14,1997. The staff concluded in its assessment that Combustion

Engineering and the affected licensees have taken appropriate steps to ensure adequate

integrity of Combustion Engineering designed weld sleeves. The inspectors verified that

licensee commitments to reinspect the 28 welded sleeves (which exhibited eddy current

indications during the post-installation examination) were accomplished using the plus

point probe and revised nondestructive examination inspection criteria during Refueling

Outage 2R12 (1997), with 1 of the 28 sleeved tubes plugged as a result of the 1997

examinatims. The inspectors ascertained that the licensee did not plan on installing

welded sleeves during the next Units 1 and 2 refueling outages (i.e.,1R14 and 2R13),

and had an action item, LIR L98-0002, that was due June 30,1998, to establish

acceptance criteria for dispositioning eddy current indications in sleeve welds. The

inspectors concluded that the establishment of appropriate eddy current acceptance

criteria would preclude further instances of potentially defective sleeved tubes being .

placed into service.

E8.3 (Closed) Insoection Followuo item 50-368[9714-04: _ Review of examination provisions

for two sleeved tubes with identified potential for limited service life before initiation of

primary water stress corrosion cracking.

Licensee review of this inspection followup item identified that the two sleeved tubes

'

that had been identified to have the potential for a limited service life (i.e., Tube -

, Row 10/Line 108 and Tube Row 72/Line 118 in Steam Generator A) had been

previously plugged and, thus, examination provisions were not applicable. Specifically,

"

' Row 10/Line 108 was removed from service during Refueling Outage 2R12 as a result

of the identification of a single axialindication at the 01H eggerate support. Tube

,

i

y

f

n-

y.

- .-

m

-13-

Row 72/Line 118 was found to have been removed from service during the sleeve

installation outage (Refueling Outage 2R11) because of the identification by eddy current

examination of the presence of a blowhole in the sleeve weld. The inspectors confirmed

that the tubes had been removed from service by review of the steam generator repair

history for the two refueling outages.

E8.4

(Closed) Licensee Event Reoort 50-368/2-97-008: Hindsight review of Outage 2F96-1,

1 steam generator eddy current data indicated that bobbin coil distorted support indications

,

. were not dispositioned for further charactenzation, resulting in potentially degraded tubes

~

remaining in service for approximately 5 months.

The results of inspection followup of this licensee event report are documented in .

Section E8.1 above.

E8.5 ;(Closed) Unresolved item 50-368/9628-02: Use of motorized rotating pancake coil eddy

current data to override previously acquired bobbin coil data, which exhibited steam

generator tube defect indications in excess of technical specification repair limits.

During Outage 2F96-1, the inspector noted an instance where a tube in Steam

Generator A (i.e., Tube Row 40/Line 46) was planned to be left in service despite the

identification during bobbin coil data analysis of the presence of a 45 percent through-

wall defect in the sludge pile region. Section 4.4.5.1.7 of the Unit 2 technical

specifications establishes a 40 percent through-wall plugging or repair limit. .The

licensee representative stated that this determination was made as a result of a

.

subsequent motorized rotating pancake coil examination indicating no flaw was present

I

at this location. Licensee review identified that there was a total of five tubes which

exhibited free span bobbin coil signals that corresponded to a through-wall range of 40-

,

53 percent, and for which subsequent motorized rotating pancake coil examinations

indicated no defect was present. The licensee was informed that this matter was

considered a compliance issue and that the appropriateness of disregarding rejectable

,

bobbin coil values was considered an unresolved item pending review by the Office of

Nuclear Reactor Regulation. The licensee elected to plug the five tubes with the free

span bobbin coil indications prior to returning Unit 2 to service.

.

Review by Office of Nuclear Reactor Regulation staff concluded that the practice of

- dispositioning indications detected with a bobbin coil probe (and depth sized greater than

the repair limit) via confirmation with'a motorized rotating pancake coil probe may be

<

acceptable.: This review noted the inherent susceptibility of the bobbin coil to interfering

/ conditions (e.g.,'in addition to flaws, changes in tube geometry with respect to the coil,

iconductive deposits; tube electrical and mechanical properties, and the presence of

structures _such as tube stpport plates within the magnetic field can all affect the eddy

_

-

l current signal response). Because the bobbin coilis sensitive to any of these conditions,

.

,

V

~and coupled with the fact that a large volume of material is interrogated by the coil at any

.

.

moment in time, signals generated from the bobbin probes could be a response to actual

. tube degradation, or a combination of the above factors. The acceptability of the

J

'

r

h

b

g.

4

,

's

.

4 [( -

.j

'

'

--

-

,

.

.

-14-

licensee approach was indicated by the Office of Naclear Reactor Regulation staff to be

dependent on factors such as: (1) the inability of the bobbin coil to adequately depth size

the mode of degradation, (2) the ability to demonstrate that the motorized rotating

pancake coil probe has a threshold of detection approximately equal to or less than the

repair limit in the technical specifications for the mode of degradation of interest, and

(3) consideration of allinformation provided from the bobbin coil examinations (e.g.,

phase rotation, signal-to-noise ratio).

The inspectors noted that a number of steam generator eddy current program

enhancements and changes were initially implemented by the licensee in Refueling

Outage 2R12. The changes included the increased use of confirmatory motorized

rotating pancake coil probe examinations for further diagnosis of bobbin coil signals,

with confirmation of the presence of a flaw resulting in removal of the tube from service

regardless of estimated flaw size. Overall, the inspectors considered the licensee

actions to be appropriate, particularly with respect to tube flaws present at eggerate

locations. Some axial flaws present at eggerate locations were noted by the NRC

consultant during Refueling Outage 2R12 (see Section M1.1, NRC inspection

Report 50-313/97-14; 50-368/97-14) as not being routinely identifiable if bobbin coil data

analysis was cursory. Analysis of the bobbin coil data was made more difficult by the

small amplitude of many of the flaws and the limited ability to eliminate the effects of

eggerate structures on the eddy current signal response. The inspectors concluded that

the introduction in Refueling Outage 2R12 of greater conservatism and rigor in analysis

of bobbin coil data, coupled with increased use of the motorized rotating pancake coil

examination technique for further characterization of bobbin coilindications, provided

greater assurance that small axial tube flaws at eggerate locations would be detected

and removed from service.

The NRC consultant considered the threshold of detection of the 0.115-inch pancake coil

to be significantly below 40 percent through-wall for a free span flaw. Licensee staff

were questioned regarding the threshold of detection of the motorized rotating pancake

coil probe that was indicated by the laboratory examination results from tube pull

specimens. The information provided in response by licensee staff, from the results of

'

laboratory examination of tubes pulled in 1992, included an example of a false call by a

bobbin coil. Tube Row 19/Line 55 in Steam Generator B was identified by the bobbin

coil to contain a flaw (estimated to be 31 percent through-wall) in the sludge pile region of

the steam generator. Motorized rotating pancake coil examination of the tube at this

location indicated no degradation was present. Subsequent metallographic examination

'

of the removed tube confirmed no degradation was present. Bobbin coil examination of

Tube Row 19/Line 55 also produced a distorted indication at the 01H eggerate support.

Confirmatory motor!:ed rotating pancake coil examination identified that an axial flaw

was prescrit, with an estimated length of 0.72 inches and average through-wall depth of

46 percent. The actual flaw depth at the 01H eggerate support location was measured in

the laboratory at 25 milincrements along the flaw length. The available data in the report

(TT-MCC-210, Volume 1) showed through-wall depth measurements for a 0.5-inch

I

portion of the flaw, which ranged from 8-52 percent. Similar results were obtained for

-_-

3

.

..

!

-15-

'

'

Tube Row 96/Line 116 in Steam Generator B. Bobbin coil examination detected a flaw

. indication at the 02H eggerate support, with an estimated through-wall depth of

41 percent.- Motorized rotating pancake coil examination showed that an axial flaw was

L

present, with an estimated length 0.51 inches and average through-wall depth of

39 percent. Laboratory examination at 25 mil increments along the flaw showed that the '

through-wall depth ranged from 29-59 percent.

-

,

. The inspectors considered that the length of flaw detected by the motorized rotating

pancake coil, for these two pulled tubes, was an indicator that the depth threshold of

detection at eggerate locations was below the 40 percent through-wall plugging limit of

the technical specifications. The inspectors concluded that the current diagnostic use of

motorized rotating pancake coil examinations (for confirmation of the presence of flaws

at eggcrate and free span locations), with removal of all confirmed flaw indications, was

,

consistent with the requirements of the technical specifications.

V. Management Meetings

- X1

Exit Meeting Summary

The inspectors presented the inspection results to members of licensee management at

the conclusion of the onsite inspection on March 6,1998. The licensee personnel

. acknowledged the findings presented. Licensee personnel were asked whether any

,

materials examined during the inspection should be considered proprietary.' No

proprietary information was identified. Additional in-office review of the inspection

findings was performed subsequent to the onsite insoection. A second exit meeting was

conducted telephonically on March 11,1998, to inform the licensee that two potential

violations were identified during the additional review.

.

Mi A, -

.-.g

' I. '

.1

y

,

+

t

.

' f,

'

,

ATTACHMENT-

4

SUPPLEMENTAL INFORMATION

r.

PARTIAL LIST OF PERSONS CONTACTED

beensee

' O. Anderson, Plant Manager, Unit 2

' O. Ashley, Supervisor, Licensing

'0. Denton, Director, Support

. D. Harrison, Supervisor, Engineering Programs

R. Hutchinson, Vice President, Nuclear Operations

W. James, Outage Manager, Unit 2

R. Lane, Director, Design Engineering

D. Meatheany, Engineer, Engineering Programs

K. Panther, Level ill Non-Destructive Examination Examiner

S. Pyle, Licensing Specialist

R.' Rispoli, Supervisor, Engineering Programs

M. Smith, Manager, Engineering Programs

. J. Vandergriff, Director, Quality

NRC

K. Kennedy, Senior Resident inspector

INSPECTION PROCEDURES USED

IP 73753

Inservice inspection

IP 92700

Onsite Followup of Written Reports of Nonroutine Events at Power Reactor

Facilities

IP 92902

Followup - Maintenance

IP 92903

Followup - Engineering

ITEMS OPENED, CLOSED, AND DISCUSSED

Ooened

50-368/9812-02 VIO -

Inappropriate setup guidance given for analysis of plus point probe

eddy current data (Section E8.1)

50-368/9812-03 VIO

Change made to guidance in engineering standard without either

use of a standard change notice form or revision of the engineering

standard (Section E8.1)

,-

+

.

.

-2-

Closed

50-313,-

VIO

Failure to include certain ASME Code, safety-related valves in

368/9713-01

inservice test program, and failure to appropriately test certain valves

in their safety function position (Section M8.1)

50-313/9713-02 DEV Failure to meet commitments regarding inservice inspection

frequency with no subsequent notification made to NRC (Section

M8.2)

50-368/9714-03 .VIO

Lack of prompt corrective action in 1995 prior to returning potentially

defective sleeved tubes to service (Section E8.2)

50-368/9714-04 (Fl

Review of examination provisions for two sleeved tubes with

4

identified potential for limited service life before initiation of primary

water stress corrosion cracking (Section E8.3)

50-368/2-97-

LER

Hindsight review of Outage 2F96-1 steam generator eddy current

008

data indicated that bobbin coil distorted support indications were not

dispositioned for further characterization, resulting in potentially

degraded tubes remaining in service for approximately 5 months

(Section E8.4)

50-368/9628-02 URI

Use of motorized rotating pancake coil eddy current data to override

previously acquired bobbin coil data which exhibited steam

generator tube defect indications in excess of technical specification

repair limits (Section E8.5)

Ooened and

Closed

50-313/9812-01

NCV Failure to comply with Section XI ASME Code examination

requirements (Section M3.3)

f

'

\\

.

f

3-

Discussed

50-313;

VIO

(1) Unit 1 potential for steam generator tubes to be left in service that

368/9714-01

exceeded the plugging limit of the technical specifications, and (2)

Unit 2 tubes that had defects in excess of the plugging limit of the

technical specifications (Section E8.1)

DOCUMENTS REVIEWED

- Prooram Documents

Document A4.106,'" Steam Generator Tube Integrity Program," dated January 12,1998

Nuclear Energy institute Document NEl 97-06, " Steam Generator Program Guidelines," dated

December 1997

Inservice Inspection Plan Arkansas Nuclear One Unit 1, Third interval, Revision 0

inservice Inspection Plan Arkenc:c Nuclear Gno Unit 2, Second Interval, Revision 4

Schedule of 1R14 ISI Outage inspections, Rwh i 4

1

Procedures

Engineering Report 98-R-2002-01, "2P98 Mid-C. d Ntags Eddy Current Examination

Technique Qualification," Revision 0

Report ER 974855-E201," Steam Generator Pi9-Outage Degradation Assessment and Repair

Criteria for 2P98," Revision 0

Report ER-974854-E101," Steam Generator Pre-Outage Degradation Assessment and Repair

{

Criteria for 1R14," Revision 0

Procedure 5000.018 " Steam Generator Integrity Program Administration," Revision 0

Procedure 5120.509, " Steam Generator Inservice inspection implementation Program,"

Revision 0

Procedure 5120.500, " Steam Generator Integrity Program Implementation," Revision 6

ANO-2-OTH-ESP-SGMAN," Arkansas Nuclear One-Unit 2 Steam Generator Eddy Current

Training Manual," Revision 2

i

m

_p-

i

O

,

-4--

Engineering Standard HES-28. "ANO-2 Steam Generator, ECT Data Analysis' Guidelines,"

Revision 8 '

Engineering Standard HES-72, "ANO Eddy Current Data Acquisition," Revision 0

2R12 Outace Inservice insoection Examinatlons01-089

01-S-045

02 B-082

33-012

- 01 L-037 -

01-S-046

02-B-083

33-013

01-N-037

01-S-047

02-8-084

41-040

01-N-054

01-S-048

03-022

48-008

- 01-S-037

01-S-049

17-002

48-009

01-S-038

01-S-050

17-003

72-082W

- 01-S-039-

01-S-051

19-038

72-083

01-S-040 '

01-S-052

19-040

72-180

.01-S-041

01-S-053

21-064

-80-619

- 01-S-042'

01-S-054

33-005

80-629

01-S-043

01-W-037

33-006

85-401

01-S-044

02-001

1

.

.

...a..