IR 05000301/1989028

Forwards Safety Insp Rept 50-301/89-28 on 891003-1108.No Violations Noted

ML19332E331
Person / Time
Site:	Point Beach
Issue date:	11/28/1989
From:	Cooper R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Fay C WISCONSIN ELECTRIC POWER CO.
Shared Package
ML19332E332	List: IR 05000301/1989028
References
NUDOCS 8912070091
Download: ML19332E331 (2)

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NOV 281989

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Docket No. 50-301 Wisconsin Electric Power. Company ATTN:

Mr.'C. W. Fay L

Vice President b

Nuclear. Power-231 West Michigan Street - P379 Milwaukee, WI 53D1 m

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Gentlemen:

This refers to the routine safety inspection conducted by Hro J. F. Schapker of.this office on October 3 through November 8,~1989, of inservice ins)ection (ISI) activities at the Point-Beach Nuclear Power Statica, Unit 2, aut1orized by NRC Operating License No. DPR-27 and to the discussion of our findings with

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Mr. J. J. Zach at the conclusion of the inspection.

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'The enclosed copy of our inspection re) ort identifies areas examined during the. inspection. Within these areas, t1e inspection consisted of a selective

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examination of procedures and representative records, observations, and interviews with personnel.-

No violations of NRC requirements were identified during the course of this inspection.

In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC Public Document-Room.

'We'will gladly discuss any questions you have concerning this inspection,

Sincerely,

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R. W. Cooper, II, Chief

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Engineering Branch

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Enclosure:

Inspection

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Report No.. 50-301/89028(DRS)

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Wisconsin Electric Power Company

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REGION III'

y Report No.:

50-301/89028(DRS)

e Docket No.:.50-301 License'No.: DPR-27 Licensee:

Wisconsin Electric Power Company 231 West Michigan Street - P379

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Milwaukee, WI 53201 Facility Name: Point Beach Nuclear Power Station - Unit 2

. Inspection At: Two Rivers, WI

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. Inspection Con ucted:

ctober 3-5, 16-18, 30-31, and November 8, 1989 V

///27/#9 Inspector:

.'F Schap r Date '

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Approved By:

D. H. Danielson, Chief

"h7/(f Materials and Processes Section Date

. Inspection' Sumary Inspection on October 3-5, 16-18, 30-31, and November 8, 1989 (Report No.

50-30VB90ZBiDR5))

Areas Inspected: Routine _ unannounced inspection of inservice inspection (ISI)

activities including review of program (73051), procedures (73052), observation of work and work activities (73053), data review and evaluation (73755); of unresolved items (92701); of the ultrasonic examination (UT) of the steam

' generator transition field weld (92704), reactor vessel volumetric examination (73753), and observation of plug repairs in response to NRC Bulletin 89-01 (92703).

Results: Of the areas inspected, no violations or deviations were identified.

Eiring the course of the inspection, the following was noted:

Based on the areas reviewed, the licensee appears to have an effective

ISI Program. The licensee contracts for ISI services and performs ISI activities to ASME-Section XI requirements, utilizing state of the art ISI equipment and procedures.

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The. licensee performed volumetric examination (ultrasonic examination) of

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the reactor vessel welds; an indication in excess of.ASME Section XI Code

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requirements was disclosed utilizing standard beam spread flaw sizing i

techniques.

(The licensee experienced similar flaw indications in Unit 1

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while performing this examination during the previous refueling outage

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for Unit 1.) Consequently, the licensee formed a joint venture with

F another licensee and employed the services of Southwest Research Institute

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(SWRI) to develop ultrasonic sizing techniques for these type of

indications. Details of this technique are included in this report.

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The licensee expanded their ISI program to inspect the steam generator

transition cone weld due to industry concerns of cracks developing in this weld.

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DETAILS

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Persons Contacted

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Wisconsin Electric Power Company (WEPCo)

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• J. Zach, Plant Manager

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• G. Sherwood ISI Engineer

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• J. Kohlway, Nuclear Engineer K. Crowly, Nuclear Engineer

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• J. Knorr, Regulatory Engineer

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G. Maxfield, General Superintendent, Operations

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R. Winjet, Nuclear Engineer

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J. Michaelson, Quality Assurance

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EBASCO Services Inc. (EBASCO)

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J. O'Neil Level II UT l

P. Deeds, Level II UT

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Southwest Research Institute (SWRI)

H. Diaz, Level III UT P. Gaines, Level II UT i

J.'Alejandro, Level II UT

Hartford Steam Boiler Engineering and Insurance Company (HSB)

D. Oakley, ANII Westinghouse Electric Corporation (W)

'J. Murphy, Quality Control J. Marburgur, Quality Assurance Engineer U. S. Nuclear Regulatory Commission (V. S. NRC)

• C. Vanderniet, Senior Resident Inspector The inspector also contacted and interviewed other licensee and contractor

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employees, b

• Denotes those attending the exit meeting on November 8, 19B9.

2.

Licensee Action on NRC Bulletins (0 pen) NRC Bulletin 89-01: Failure of Westinghouse Steam Generator Tube

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. Mechanical Plugs.

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' Background f

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Numerous plants have experienced primary water stress corrosion cracking (PWSCC) and leaks of Westinghouse mechanical plugs.

On February 25, 1989,

North Anna, Unit 1 experienced a mechanical plug failure followina a

reactor trip during a feedwater isolation transient. The plug failure caused a 75-gallon per minute (gpm) primary-to-secondary leak and was the

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subject of NRC Information Notice No. 89-33, " Potential Failure of

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Westinghouse Steam Generator Tube Mechanical Plugs." The failure mechanism

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involved a full circumferential severance of.the top portion of the plug from the body of the plug.

The top portion of the plug was propelled up

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g the length of the affected tube by primary system pressure to a

)oint just

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above the u-bend tangent )oint where it impacted and punctured. tie outer o

curvature of the tube.. Tie top portion of the plug subsequently impacted and dented an adjacent tube. The failed plug was installed in November 1985.

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Licensee Action

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L Steam generator maintenance records were reviewed to identify the installation date, location, and heat number for all installed mechanical

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plugs. Using the methodology of WCAP-12244, Revision 1, and the benchmark degradation rate for Nilstone 2, estimated plug lifetimes were determined for the susceptible heats identified by Westinghouse. For Unit 2, both

.the 7/8 inch mechanical plug and the 3/4 inch sleeve mechanical plug are installed. Plug lifetime estimates are determined for each plug size as a function of the hot leg and cold leg operating temperatures. Based on

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this estimated plug lifetime Unit 2 susceptible hot leg plugs were repaired during this outage. The licensee utilized the plug-in-plug (PIP)

process._to. correct the deficient plugs.

Susceptible cold leg plugs will be repaired during a future outage as lifetime expectancy calculationa exceed 3,000 effective full power days i

of operation prior to plug failure.

Inspection x

The NRC inspector observed installation of the PIP for Unit 2 Steam l

Generators (SGs) "A" and "B".

A total of 189 repairs were made to the l

IGSCC susceptible plugs (65 on SG "A" and 124 on SG "B").

One sleeve plug in the 'yG was replugged due to failure of PIP to insert properly.

l The NRC inspectvr reviewed installation, welding, and inspection procedures associated with the PIP installation, observed welding, and visual inspection of completed welds. The PIP installation was performed utilizing robotic equipment to reduce the man-rem exposure.

This bulletin remains open pending future repairs to Unit 2 cold leg susceptible plugs.

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3.

Licensee Action on Previous Inspection Findings

(Closed) Unresolved Item (301/87021-01(DRS)): Reactor Coolant Loop Piping e

Long Seams Not Included in ISI Program.

The licensee previously discovered evidence which indicated certain portions of the reactor coolant piping were fabricated with welded long

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t seams, which had not been previously examined in accordance with ASME

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Section XI Code requirements, r

The licensee has confirmed the existence of the welded long seams and e

additional girth welds (4) not previously examined as required by K

Section XI. The exclusion of these welds resulted from inaccurate as-built t

drawing records where some shop welds were not recorded, thereby being

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F excluded from the Inservice Inspection (ISI) Plan. The licensee has implemented the following corrective action:

(1) all welds not previously

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inspected have been examined in accordance with ASME Section XI Code l~

requirements; (2) a nonconformance report has been issued to address the-inaccuracy of the ISI plan due to the as-built drawing discrepancies; (3)

f a review of sho) drawings and construction documentation to verify the-accuracy of as-)uilt_ drawings is being performed to assure other systems'

ISI-programs are accurate; and (4) examination of the previously unidentified longitudinal welds utilizing special ultrasonic equipment and procedures developed for cast stainless steel has been performed with no recordable indications. The licensee's corrective action was appropriate to assure ISI program deficiencies are identified and corrected.

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Inservice Inspection (ISI) Unit 2 a.

General The_NRC inspector reviewed the second ten-year Inservice Inspection Plan for the third period, second outage. The Inservice Inspection Plan conforms to the American Society of Mechanical Engineers (ASME),

Boiler and Pressure Vessel Code,Section XI, 1977 Edition, Summer 1979 Addenda. A relief request in accordance with 10 CFR 50.55a(g)

(5)(iv) had been applied for and accepted by NRR and for those inservice examinations to be performed during this period. The services of an Authorized Nuclear Inservice Inspector (ANII) were arocured and the ISI procedures and personnel certifications have

)een reviewed by the ANII.

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The licensee contracted SWRI to perform reactor vessel mechanized

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ultrasonic examination; EBASCO and WEPCo QA performed the remainder of the ISI in accordance with ASME Section XI 1977 Edition, Summer 1979 Addenda. Westinghouse Electric Company (W) performed eddy current examination of the steam generator tubes.

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ISI: Documentation Review. '

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--The NRC inspector reviewed documents relating to the following:

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-Eddy-current equipment calibration.

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-U\\trasonic(instruments, calibration blocks, transducers and.

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ultrasoric examination couplant certifications.

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tiquid penetrant material certifications.

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y a0E personnel certifications in accordance with SNT-TC-1A.

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NDE procedures utilized for ISI.

• NDE'chlibrationiand examination reports.

• Eddy current-examination data reports.

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No' violations or' deviations were identified.

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Observation of Work Activities

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.The NRC inspector observed ~ work and discussed examinations with NDE examiners. These activities included observation of calibrations

and" examination performance, and review of documentation of the

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.following NDE:

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Hydrostatic test of spray additive tank.

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Magnetic' particle examination.(MT) of reactor coolant pump

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-studs.

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Ultrasonic examination (UT) of steam generator (SG) transition

cone girth weld.

-Volumetric examination of reactor vessel (RV) welds.

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Eddy current exam N tion (Ei) of SG tubes.

• Visual examination of the RV core barrel storage fixture.

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Work activities were performed in accordance with approved

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procedures, utilizing calibrated nondestructive examination (NDE)

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equipment.. Detection of and resolution of flaws disclosed by NDE

procedures are discussed in Paragraphs 5, 6, and 7 of this report.

Ultrasonic Examination of Steam Generator (SG) Transition Girth Weld

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Tne licensee performed ultrasonic examination of the upper shell to

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transition cone girth weld on SG "B".

This examination was not included as:part of the scheduled ISI but was perforraed due to industry identified Lproblems'of cracking in this weld (Indian Pi nt, Surry).

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The. examination'was performed utilizing pulse echo type UT detection

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i instruments with A-Scan presentation.

Piezoelectric transducers at

. nominal beau angles of zero degrees (longitudinal), 45' and 60' shear wave

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(transverse) were used.= The UT procedure complied with the requirements

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specified in ASME Section XI, 1977 Edition, Summer 1979 Addenda.

Twdre)ortableindicationsweredetected. One indication had no measureble B>

througi wall dimension, the other was subsurface, sized at 0.55 inch long

wall vs. 4.3%gh' wall. This is within ASME Code allowable (4.0% through

by 0.30 throu E

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through wall Code maximum).

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-The NRC inspector observed the sizing of the referenced indication and

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scanning of the weld and adjacent base metal at 20 db hot, to identify any ID surface indications as found'at Indian Point and Surry, rio relevant

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indications were observed.

f No violations or deviations were identified.

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Reactor-Vessel' Volumetric Examination

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The NRC inspector reviewed the ISI plan to perform mechanized ultrasonic examinations of' selected areas of the reactor pressure vessel and adjacent

piping and remote visual examinations of selected reactor pressure vessel internals and interior surfaces of the reactor vessel.

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The following Southwest!Research Institute (SWRI) procedure was reviewed:

NDT-700-11 "Hechanized Ultrasonic Inside Surface Examination of-Ferritic Vessels Greater Than 2.5 Inches in Thickness", Revision 11.

The NRC-inspector reviewed the qualification certifications of the SWRI

ultrasonic and visual examination personnel. Certifications verified the

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-SWRI~ inspectors were qualified to SNT-TC-1A as required by applicable Code requirements.

Ths NRC inspector _ observed the UT mechanized scanning of the reactor l

vessel circumferential welds and nozzle-to-shell welds, data acquisition, w.

and data evaluation in progress.

Indications in excess of Code allowable

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were' detected'in the safety-injection nozzle to reactor vessel weld l-

(RPV-97-01-B).

Initial sizin, was performed using ASME Code sizing technique'.

'SWRI developed flaw-sizing ultrasonic techniques to be used in nozzle configurations present in the Point Beach and R. E. Ginna reactor pressure vessels (RPV). RPV UT indications disclosed in previous examinations were sized'in excess of Code allewables at these sites. Subsequent fracture mechanics analysis was performed which met Code allowables for continued operation with augmented examination of the affected welds. The SWRI L

special-sizing technique utilized focus beam transducers and time of flight sizing. The method of sizing flaws was demonstrated to the WC utilizing test mockups of the same material and geometric configuratluns

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as the RV nozzles. Flaws consisting of machined notches and flat-bottom

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holes were placed at locations representing the nozzle-to-shell weld

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fusion'11ne nearest the nozzle bore.

In addition, Electric Power Research

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s-Institute (EPRI)-mockups of two safety injection nozzles, one containing

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' artificial flaws'and the other containing actual flaws, were used This

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method,of sizing was demonstrated to be much more accurate in flaw sizing

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than-the ASME. Code sizing.

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Consequently, the focus beam / time of flight. sizing technique was used to i

evaluate the flaws in the Unit 2 Point Beach reactor vessel-safety

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injection nozzle.' Results of the flaw sizing concluded that the flaws

'did,notl exceed Code allowables.

The NRC inspector requested'that flaw sizing data from the previous reactor. vessel examination performed in 1977 be compared to the present

. flaw using the. flaw sizing method (ASME Code) utilized in 1977. This

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w comparison was ne.assary to assure the flaw as sized in 1977 had not increasedLin size. The ultrasonic sizing comparison to the 1974 ASME Code

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(50% DAC) was utilized and concluded that the flaw had not changed and

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therefore -is not service induced but a fabrication abnormality such as welding slag. The_NRC inspector also reviewed fabrication radiographs of

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-the safety' injection nozzle to reactor vessel weld. The radiographs were inconclusive 1due to age, artifacts, and inadequate radicgraph reader sketches. Some slag indications were present but location could not be discerned as the sketch reference point was not obvious. The NRC inspector contacted NRR/EMTB, who concurred with the-acceptability of the SWRI b

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sizing technique. This technique was previously used at Ginna and a 1 Safety Evaluation Report (SER);was issued by NRR.

The licensee's reactor-vessel volumetric examination was adequate with

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r_espect to. meeting the safety objectives of the ASME Code requirements.

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-Eddy Current Examination of Steam Generators

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The licensee employed (W) to perform eddy current examinations of the steam generator tubes as required by Technical Specifications (TS).

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utilized the MIZl18 multi-frequency acquisition and analyzer system to conduct the' examinations. The "A" steam generator hot leg inspection results indicated four'(4) tuoes degraded equal to or greater than 40%

of the wall thickness, one tube with an undefined signal, twenty (20)

-tubes with axial indications in the tube sheet area, and two (2)

restricted ~ tubes. All of these tubes were plugged.

In the "B" steam generator, a total of'seven (7) tubes were degraded = equal to or greater

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than 40% of the tube wall thickness, and two tubes had-axial indications in the tube sheet area. Six of these tubes were plugged and the

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remaining three were included in the. sleeving program.

The dedgradation of:these steam aenerator tubes is typical of industry experience.

The eddy current inspection scope consisted of the following:

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A 10% full length sample on "3" SG and a 9/61 full length sample on "A" SG (TS requires 3.0%).

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Previously degraded tubes that had not been repaired.

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_A'10% sample of the tubes which had sleeves installed in the hot

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leg to-the,first-support plate looking particularly for tubesheet

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crevice corrosion.

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100% of the remaining.unsleeved tubes in the hot leg to the first

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!suppert plate:looking particularly for tubesheet crevice corrosion.

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,f, A 10% sample of tubes which had sleeves mitalled in 1988 in the

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cold. leg in each generator.

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.100% inspection of the unsleeved tubes in the "B" SG from the cold

leg tube end to the first support plate for wastage and pitting.

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.A~14% sample of the "A" cold leg unsleeved tubes from the tube end

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.to the first support ~ plate for wastage and pitting. This included

~ ll of the peripheral tubes and those tubes in the area where a

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wastage had-been noted in previous inspections.

c The NRC. inspector observed ~the eddy current examination (ET) of the steam

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. generator. tubes.in progress, verified certifications of ET equipment, calibration standards, and reviewed the qualifications and certification of the ET: examiners.-

'The NRC inspector observed the' sleeving of SG tubes-in the Unit 2 "B"

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!SG. The-licenser elected to install mechanical. sleeves in SG-tubes'in Laccordance with iechnical Specification 15.4.2.A.6.

Selection of tubes to be-sleeved was based.on the ET. data. ;As part of the ongoing maintenance

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and. inspection programs by the-licensee, ET has been performed on the Unit 2 SG's cold leg tubes during past outages.- Results from the'most recent

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inspections, conducted during this and previous outages, indicated that a significant number of Unit 2 SG "B" tubes increased in tube wall degradation.

However, the. degradation was less than the plugging limit defined in

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Technical Specification 15.'4.2.A.5(a). =The majority of this degradation was in the cold leg.where previous tube thinning was detected.

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The: sleeving program: included both repair and preventive sleeving. Repair sleeves were; performed on those tubes which had degradation'in excess of the plugging limit but were technically capable of being sleeved.

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Preventive 1 sleeving was performed on the tubes'having ET indications which.may. exceed the plugging limit in the future.

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A total of 298 sleeves were installed in the "B" steam generator cold leg during this outage. These were installed as a pre /cntive maintenance

' measure due to tube thinning in the sludge pile region above 'he tube sheet...Three SG tubes were sleeved which exceeded the plugging limit.

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s The licensee's preventive maintenance to prevent tube degradation includes-sludge lancing,-crevice flushing prior to heat up to normal

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operating temperature and maintaining chemistry controls in accordance su with prescribed guidelines. Sludge lancing performed this outage e

resulted.in:154 pounds of sludge removal from SG "A" and 102 pounds from-

~SG "B".

Visual examination of' secondary side after lancing was

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satisfactory.-

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-The licensee's eddy. current examination ana tube degradation preventive.

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' maintenance program are adequate to assure safe operation of the steam generators.-

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Exit Interviews

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The Region III' inspector' met with licensee representatives-(denoted in

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Paragraph-1) at:the-conclusion of the inspection on November B,>1989. The

^Y inspector summarized the purpose and findings of.the inspection. The

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licensee representatives acknowledged this information. The inspector

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also' discussed the likely informational content of the inspection report

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with1 regard to documents or processes reviewed during the inspection.

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Thel licensee representatives did not identify any such documents / processes

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