ML20207N233

Insp Rept 50-219/88-25 on 880829-0902.No Violations Noted. Major Areas Inspected:Licensee Actions in Response to NRC Bulletins 79-02 & 79-14,Deficiencies Noted in Some Pipe Support & Pipe Support Base Plates

ML20207N233
Person / Time
Site:	Oyster Creek
Issue date:	10/04/1988
From:	Carrasco J, Chaudhary S, Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20207N224	List: ML20207N222 ML20207N233
References
50-219-88-25, IEB-79-02, IEB-79-14, IEB-79-2, NUDOCS 8810190039
Download: ML20207N233 (6)

See also: IR 05000219/1988025

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 50-219/88-25

Docket No. 50-219

License No. DPR-16

Licensee: GPU Nuclear Corporation

P.O. Box 388

Forked River, New Jersey 03731

Facility Name: Oyster Creek Generating Station

Inspection At: Forked River, New Jersey

Inspection Conducted: August 29 - September 2, 1988

Inspectors: _

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S. K. L;naudnary, Venior Reactor Engineer ' d te

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• J. Carrasco. React [r Engineer ' cate

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Approved by: _ _

. Strosnider, Chief, MPS, EB, DRS

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Inspe_ction Summary: Routine unannounced inspection on August 29 - September 2,

1983 [ Report No. 50-219/88-25)

Arcas__ Inspected: Adequacy of licensee actions in response to NRC Bulletins

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79 02 and 79-14; Follow-up inspection for items identified by NRC in this and

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other areas in previous NRC inspections.

Results: No violations were identified. However, deficiencies were

identified in soma pipe support and pipe support base plates. An inspection

is planned to confirm effective completion of Bulletin 79-02 and Bulletin

79-14 activities prior to restart from the upcoming 12R outage. NRC

Bulletin 78-12 was closed.

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DETAILS

1.0 Persons Contacted

General Public Utilities-Nuclear

J. Barret, Plant Operations Director

J. Barton, Deputy Director, OC

T. Corrie, Quality Control Manager

P. Dix, Manager, Technical Support (PM)

V. Foglia, Manager, Technical Functions, OC

J. Kowalski, Licensing Manager, OC

B. J. Rogers, Licensing Engineer

T. Snider, Manager, Plsnt Maintenance

T. Quintenz, Manager, Material Assessment

U.S. Nuclear Regulatory Commission

J. Wechselberger, Senior Resident Inspector

D. Lew, Resident Inspector

E. Collins, Resident Inspector

The inspectors also contacted, and were assisted by other technical,

quality control, administrative, and craft personnel, as their work

interfaced with the scope of this inspection.

2.0 The Purpose and Scope of the Insnection

The purpose of the inspection was to assess the adequacy and effectiveness

of the implementation of the licerisee's program to inspect, analyze, and

evaluate pipe supports in the plant to meet the requirements of NRC

Bulletins 79-02 and 79-14. The inspection focused on the installed

supports that were categorized as having a factor of safety of twenty (20)

or more and had not been examined or modified by the licensee. The

inspector also examined some supports which were modified and upgraded

to meet the requirements. Due to the emphasis on physical examination

and measurements, minimal documentation review was performed, except for

drawings and records of visual examinations performed by the licensee in

response to NRC Bulletin 79-14. These records were reviewed to determine

the location and identification of the selected supports for inspection.

3.0 Re- ww Criteria

> *1a+ ins 79-02 and 79-14 were used primarily to define inspection

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rr- C The previous NRC inspection reports were used to define

ti i ....e commitments and any necessary followup inspection to assess

the status of cpen items related to these bulletins.

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4.0 Verification of "As-Built" Condition of Pipe Supports

The inspector selected a sample of 20 pipe supports to verify their

actual configuration, technical adequacy of their installation, and

quality of workmanship. The support configuration was visually examined

and compared to the applicable design drawing of the selected support. In

addition, for five selected supports the existing torque values for the

concrete expansion anchors used in the supports were verified. The torque

was verified by a crew of maintenance mechanics and helpers provided by

the licensee using a calibrated torque wrench. The torque verification

process also was witnessed by the licensee's QC inspector assigned to the

verification crew.

The licensee had informed the NRC that a large number of supports in the

plant were designed very conservatively; thus, indicated a large factor

of safety (20 or more) during their anaiysis and evaluation. These supports,

therefore, were excluded from any detailed inspection or modification by

the licensee. The inspector selected a number of supports from this group

I to determine the validity of licensee's conclusion, because, the analytical

i determination of a f actor of safety is dependent on the correct and proper

( installation of the support as designed. Additionally, some supports that

l were recently modified were included in the sample to ascertain the

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adequacy of installation.

Findings

Based on the above examinations, discussion with engineers, and review of

pertinent documents, the inspector rade the following observations.

1. On plate 'C' of support no. 212-BP-NZ-2-Rl-84, the nut on the lower

right corner appeared to have been tightened. This was apparent due

to the breaking of the paint covering the nut and the base plate.

This plate was not included in the group of supports needing modifi-

cation or work. No documentation or personal recollection was

readily available to explain why or by whom work was performed on

the concrete expansion anchor.

2. In the base plate for support no. 212-47 the inspector verified the

existing torque in the expansion anchors. There were no records of

work on this base plate or its anchors since original installation.

Although the anchor nuts on the base plate showed existing torque to

be 23 FT-LB, one of the nuts was not in contact with the plate.

There was a gap of 1/16" between the plate and the nut. The nut also

indicated evidence of having been tightened although no records were

available indicating when or why the tightening was perform?d.

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3. The supports in the torus room and CR0 room had the nuts on the

concrete anchors tack welded to the base plate and the bolt was cut

flush with the top of the nut. Also, the grout under the plate was

excessive (over 2") in thickness. There was not evidence, documented

or visual, that the existence or absence of levelling nut under the

plate had been verified. The drawings for similar support in QC file

indicated no requirement of tack weld and a grout thickness of one

inch. The tack welding and cutting of anchor bolts makes it difficult

to verify the length of the installed anchor bolt and impossible to

check the torque valves of the installed anchor bolts.

4. Two supports in the Liquid Poison Sy . tem which had been modified

recently, were examined for conformance to design configuration and

existing torque in the anchors. The inspector observed that one of

the new "Ram-set" bolts installed in one support appeared to show an

imbediment depth of only 11/16". Another bolt in the other support

indicated an existing torqua value of 11Ft-LB. The initially

required installatiot ,orque was 20 Ft-LB. This indicates an

uncommonly large loss of approximately 45*. of the torque in less than

a year due to relaxation or possibly deficient original installation.

5. A previous NRC inspection (IR 85-14) identified that hot and cold

hanger settings on spring can hangers and supports were not verifi-

able due to paint over the indicator plate. The inspector observed

that similar conditions existed on the three spring hangers selected

during this inspection. Although, on two of the three spring cans

hot settings could be guessed, they were not sufficiently clear to

accurately establish the setting with reasonable confidence.

5.0 Drawing Systems

In the course of reviewing drawings for support configurations, the

inspector made some general observations about the licensee's plant design

drawing system. The observations are as follows:

1. Several dif ferent isometric drawings are available, and are used by

different groups on site. The inspector had considerable difficulty

in establishing which of the drawings were the current applicable

drawing for selected systems. Also, the system used to identify

support. is different on the various drawings. For example, there

are original Burns and Roe drawings, Bergen-Patterson support draw-

ings, General Physics drawings, and in some cases drawings by other

organizations without a unified system of support identification.

The licensee has tabulated a cross-reference chart and is currently

trying to establish a unified support numbering system, but the

system is not widely available or used by dif ferent organizations

on-site.,

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2. New modifications to the systems are documented on a new drawing

series which is not clearly referenced in the applicable master

drawing of the system, making it cumbersome and difficult to

establish the current "as-built" configuration of the system.

The above observations by the inspector were brought to the attention of

the licensee management for their evaluation.

6.0 Status of Open Items

Following items remain open. These items are related to Bulletins 79-02

and 79-14.

(0 pen) yiolation, 85-14-01: Activities performed for IE Bulletins

79-02 and 79-14 were not covered by documented procedures.

(0 pen) Unresolved Itet, 85-14-02: The licensee could not retrieve

inspection checklist, marked-up drawings, and/or other field

inspection documentation during the NRC inspection.

(0 pen) Deviation, 85-14-03: Five of twelve seismic Category I systems

were not tested for anchor bolt acceptability

(0 pen) Deviation, 85-14-04.: No specific design documents applicable

to the seismic evaluations of safety-related piping were

available.

(0 pen) Unresolved Item, 85-14-05: Adequacy of the baseplate and bolt

evaluation for support NC-Z-HZ1 could not be determined.

(0 pen) Unresolved Items, 85-14-06: A verification of engineering

disposition for calculation number 8.31.208 was not available.

(0 pen) Unresolved Item, 85-14-07: No documentation was available to

verify the conservatism of the seismic span tables.

(0 pen) Unr3 solved Item, 86-14-08: Pipe support analyses of supnorts

with"Trictional loads did not include these loads in the

support calculation.

(0 pen) Unresolved Item,_8_5-14-09: This item pertained to many

deficiencies found in "as-installed" configurations of pipe

supports.

In response to NRC letter of November 7,1985, the licensee had com.nitted

to resolve the above open items by the end of 11R outage; however, the

schedule was extended to 12R outage currently scheduled for October 14,

1988. A NRC inspection has been planned to verify the resolution of the

above open items during 12R outage before the start of cycle 13.

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(Closed) NRC Bulletin 78-12: This bulletin pertained to the use of

atypical weld material in reactor vessel welds.

The licensee's NSSS supplier (GE) and reactor vessel supplier (CE)

investigated the possibility of atypical weld material use in the

reactor vessel of Oyster Creek. ~he vendor certified to the

licensee that no atypical materi&ls had been used in weldments in

the reactor vessel. This item is closed.

7.0 Management Meetings

Licensee management was informed of the scope and purpose of D.e

inspection at the entrance interview on August 30, 1938. The fi' dings of

the inspection were discussed with licensee reprace.otatives during the

course of the inspection and presented to licensee management at the

September 2,1988 exit interview (see paragraph I for attendees).

At no time during the inspection was written material provided to the

licensee by the inspector. The licensee did not indicate that proprietary

information was involved within the scope of this inspection.

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