ML20206J201

Insp Rept 50-219/87-07 on 870217-20.Violations Noted:Failure to Update Drawing to Show Main Steam Drain Containment Isolation Mod That Installed Two Blind Spectacle Flanges

ML20206J201
Person / Time
Site:	Oyster Creek
Issue date:	03/31/1987
From:	Gregg H, Kamal Manoly, Strosnider J, Varela A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20206J174	List: ML20206J167 ML20206J182 ML20206J201
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.16, TASK-TM 50-219-87-07, 50-219-87-7, NUDOCS 8704160039
Download: ML20206J201 (6)

See also: IR 05000219/1987007

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

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Report No. 50-219/87-07 *

Docket No. 50-219

License No. . OpR-16 Priority Category C

Licensee: GPU Nuclear Corporation

Oyster Creek Nuclear Generating Station

P.O. Box 388

Forked River, New Jersey 08731

Factitty Name: Oyster Creek Nuclear Generating Station

Inspection At: Forked River, New Jersey

Inspection Conducted: February 17-20, 1987

Inspectors u 44 4. 3 36 N

aroldI.Gredg. Lead'$eactof+gineer date

N d //W

K. Manoly, Lead Reactor Engineer

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date

/bMef &?kg$L 2/3//87

A. Varela, Lead Reactor Engineer date

Approved by: s

Strosnider, Chief

4 J/3/[6

date

aterials and Processes Section

Ing ection Summary: Inspection on February 17-20, 1987 and February 24, 1987

[ReportNo.50-M9/87-07)

Areas Inspected- This inspection was a routine unannounced inspection related

to the licensee's implementation of NUREG 0737 Item II.K.3.16 commitment to

reduction of challenges and failures of safety relief valves. The inspection

also reviewed the licensee's activitie; relating to items previously identified

by NRC. The inspection on February 24, 1987, was to follow-up on identified

concrete defects (cracking and spalling).

Results: One violation was identified. It involved a failure to up-date a

ifrawing to show the main steam drain containment isolation modification that

installed two blind spectacle flanges. The other activities relating to

adherence to NVREG 0737 Item II.K.3.16, and responses to several NRC identified

items were acceptably performed.

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, Details

1.0 Persons Contacted

1.1 GPU Nuclear Corporation (GPUN)

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K. Barnes, Licensing Engineer

C. Brookbank, Maintenance Construction Welding Engineer

• M. Budaj, Manager Plans and Programs

D. Collins, Mechanical Materiel Engineer

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L. Garibian, Civil / Structural Engrg and Design Mgr.

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C. Lefier, Project Engineer Tech Functions, Oyster Creek

L. Lohnes, Quality Control, Oyster Creek

! *B. DeMerchant, Licensing Engineer

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R. Fenti, Manager QA Modifications / Operations

*P. Fiedler, Vice President and Director, Oyster Creek

• J. Kowalski, Oyster Creek Licensing Manager

• J. Maloney, Manager Plant Materiel

P. Manning, Supervisor,0C Field Inspection

R. Newberry, Shift Technical Advisor

l "J. Rogers, Oyster Creek Licensing Engineer

• A. Rone, Plant CngineerinJ Director

• E, Scheyder, Maintenance Construction & Facilities Director

W. Stewart, Manager, Plant Operations

• J. Sullivan, Jr., Plant Operations Director

• R. Weltman, Mechanical Materiel Manager

1.2 US Nuclear Regulatory Commission (USNRC)

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• W. Bateman, Senior Resident Inspector

• Denotes those present at exit meeting.

2.0 (Closed II.K.3.16) NUREG 0737, Reduction of Challenges anc Failures of

Safety Rilief Valves (SRVs) '

Background

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NUREG 0737 requested licensees to evaluate their SRVs for the purpose of

reducing challenges and failures of these valves. The NUREG outlined

various methods of reducing challenges and failures. The licensee was

involved with the BWR Owners Group (BWROG) evaluation of the NUREG require-

ments and the general recommendations made by this group. By letter dated

October 3, 1984, the licensee submitted their specific response to NRC and

committed to accomplish the II.K.3.16 objectives through preventive

maintenance (item 4 of the BWROG recommendation). The commitment was made

to overhaul and test all 5 Dresser Electromatic Relief Valves (EMRVs)

l oach outage.

NRC's letter of October 23, 1984 accepted the licensee's commitment.

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Inspection Objective The objective of this inspection was to verify that

the licensee's commitments are being implemented.

Findings The inspector verified that all 5 Dresser EMRVs were overhauled

and tested during the recent 11R outage. Also, the inspector reviewed a

copy of Request for Project Approval for BA No. 323410 A3 Rev 1 for all 5

EMR valves (and pilot valves) to be rebuilt and tested at outage 12R.

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The inspector reviewed the in process hold / witness point checklists for

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all 5 EMRV's that were refurbished during the 11R outage. This checklist

was part of procedure No. 702.1.007 titled Electromatic Relief Valve Re-

moval, Disassembly, Repair, Reassembly and Installation. The inspector

verified that acceptance sign-offs by QC were in place at each activity

step through to completion of the work.

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! Based on the inspectors verification that the licensee's commitment is  ;

I being implemented, this item is closed.

3.0 Licensee Activities Relating to previously Identified Items i

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3.1 (Closed) NC4 Violation (85-23-01) Failure to follow procedure when i

unbackseating containment isolation valves V-16-1. The inspector

reviewed the licensee's response which described the problem cause,

, corrective action, and means to prevent recurrence. The problem was a ,

l result of failure to follow the procedure when unbackseating valve

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V-16-1. An operator attempted to stop the valve in the partially t

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closed position by opening the V-16-1 valve breaker. The operator

j did not realize that opening the breaker would cause a reactor water

cleanup system pump trip with the valve in a partially closed ,

position. The licensee's corrective action was to issue revision 3

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to Standing Order 33 which provides additional and more informative

j instructions regarding the unbackseating of specific valves. Additionally,

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a caution sign was to be placed at the breaker of valve V-16-1, the

! licensee's event report was made recuired reading for all licensed

operators and the event was reviewed with each operating shift.

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i The inspector verified that revision 3 of Standing Order 33 was in

place and that it provided necessary additional instructions. The

! inspector also verified that a permanent caution tag was placed on

the valve V-16-1 breaker panel stating, " Opening of V-16-1 breaker

{ will cause a Clean-Up Recirculation Pump Trip."

This item is closed.

! 3.2 (Closed)_InspectorFollowItem(85-29-02) Dresser Corporation

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reported protilems with their FTgure No. 3050 Y pattern diaphram

valves. The inspector verified that the licensee made a plant

modification which installed two removable blind spectacle flanges to

eliminate the function of the Dresser valves (valves V-1-106, V-1-107,

V-1-110 and V-1-111) as containment isolation valves. One blind

! spectacle flange was installed inside the drywell downstream of

j valves V-1-106 and V-1-107 and one blind spectacle flange was

installed outside the dry well in the trunion room upstream of valves <

V-1-110 and V-1-111.

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The inspector verified that use of the blind flanges as the means of

containment isolation effectively eliminates the through leakage  ;

problem associated with the Dresser Valves.

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This item is closed.  :

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3.3 Drawing Control Review

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During the inspection activity related to the Inspector Follow Item

. 85-29-02 of the preceding paragraph, the inspector reviewed the

Control Room drawings that were involved with the blind spectacle

flange piping modifications.

One of the two drawings (Burns & Roe drawing 2002 Rev 25), showed the

two blind flanges (Y-1-58 & Y-1-57). The other drawing (G.E. drawing

237E726 Rev 34), does not show either of these flanges and does not

reflect the modification.

Further review by the inspector determined that this drawing was not

identified as requiring the revision. The licensee's procedures No.

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D5-010 Checking of Drawings and No. 5000-ADM As-Built Drawings, have

requirements for assuring that all affected drawings are modified and

referenced.

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l The failure to update the drawing is a violation of 10 CFR 50, Appen-

, dix B, Criteria VI in the area of drawing control as documented in

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the enclosed Notice of Violation (50-219/87-07-01)

l 3.4 (0 pen)UnresolvedItem(219/86-30-01)

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i The item is related to identified cracks ard spalling f a safety and

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non-safety related concrete structures durdng the previous refueling

outage (IIR). The above concrete defects ':re documented in nine

Material and Nanconformance Reports (MNCR's), eight of which involved

, safety related structural concrete members in the reactor building.

Three MNCR's were evaluated and closed by engineering. They included:

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MNCR No. 86-830: Unsound concrete in an area 11" x 11h" on Wall

RA near wall R7 at elevation 23' - 6" i

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MNCR No. 86-804: Unsound concrete due to lamination (2" x 3"  :

i deep) and fine cracks on the north wall of the drywell exterior

in the vicinity of (16) penetration plates located 15" below the

ceiling of floor at elevation 23' - 6'

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MNCR No. 86-717: Deterioration (spalling and cracks) in the ped-  ;

j estals of the main condenser $ in the turbine building ,

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(non-safety).

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The remaining six MNCR's were evaluated by engineering and con-

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ditionally released. They included:

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MNCR No. 86-802: crack running east-west between colunin lines

RD & RE in the underside of the 75'-3" reactor building 3'- 0"

thick floor slab.

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MNCR No. 86-769: Cracks in girder along column line RE in the

area between the fuel pool wall and the reactor wall along col-

umn line R7 below floor slab at elevation 75'- 0" (diagonal

cracks on both sides of the beam and along the bottom side near

the reactor wall support)

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MNCR No. 86-859: Spalling in concrete beam No. 3.510 below

floor at elevation 75' - 0" near support No.CC-4-H27 approxi-

mately 12' west of column line RD along R6 line

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MNCR No. 86-753: Crack in beam No. 2B7 supporting floor at ele-

vation 51' - 0" between column 81 and beam 2833

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MNCR No. 86-814: Three spalled areas on underside of slab at

elevation 51' - 0" and hatritne cracks on beam along column line

R6 between column lines RD & RE.

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MNCR No. 86-870: Cracks and rust in three areas in the drywell

exterior wall at elevation 95' - 0".

The inspector examined accessible areas of concrete defects depicted

in the above MNCR's and discussed with licensee representatives the

proposed plan for final evaluation of defects with structural integ-

rity significance. The inspector was informed that cracks identified

in concrete members in four MNCR's (Nos. 753,789,802,814) will be

monitored using a calibrated crack monitoring device by Avongard

products. The crack monitoring will involve 33 locations on the af-

fected members. Further, the inspector reviewed the evaluation per-

formed by the licensee for cracks identified in MNCR's No. 789 & 802.

The evaluation was documented in TOR-809 and was based on results of

finite element analysis of the spent fuel pool performed in 1981.

The following observations were noted and discussed with licensee

engineering representatives:

1. Results of the analysis did not includo all shears and moments

of slab elements adjacent to the crack. Resultant moments and

shears were averaged for selected elements and compared against

member capacities (Mu & Vu). Based on the size of elements used

in the analysis, averaging of results does not have a technical

justification.

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2. Altered concrete properties, resulting from cracking due to

service and design loads and other initiating causes, were not

incorporated in the analysis. The analysis was performed using

the ANSYS computer code and utilized element STIF 63 for

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representing flat plates. This element can represent isotropic

homogeneous materials only. Uncracked (gross) concrete section

properties were employed in the analysis. The changes in resultant

forces based on cracked analysis will vary depending on the

magnitude of these forces from the uncracked analysis.

3. The evaluation was based on concrete and reinforcing steel data

from original construction drawings. Since the original con-

struction did not have QC documentation, the assumptions regard-

ing concrete strength, longitudinal and shear rebar size and

spacing should be verified for validation of the analysis

results.

4.. The proposed licensee approach for crack monitoring does not

include determination of crack depth through the thickness of

affected concrete members.

5. The conclusions from the licensee's evaluation did not provide

specific rationale for the identified slab and girder cracking.

The report referenced several potential factors as contributing

to the cracking. They included stress concentration, concrete

quality, non-uniform shrinkage, support settlement, creep and

vibration of equipment resting on the slab above the girder. No

specific justification was provided in support of any of the

above postulated causes.

The unresolved item will remain open pending licensee action and NRC

review of: observations 1 through 5 identified above, (2) final en-

gineering evaluation and disposition of all identified concrete do-

fects, and (3) implementation of the crack monitoring program pro-

posed by the licensee.

4.0 Exit Meeting

The inspector met with the licensee's representative (identified in para-

graph 1.0) at the conclusion of the inspection on February 20, 1987, to

summarize the finding of this inspection. The NRC Senior Resident Inspec-

tor, W. Bateman, was also in attendance.

The inspection of February 24, 1987, was concluded on that date and a

summary of the inspection findings was provided to the licensing repre-

sentative.

During this inspection, the inspector did not provide any written material

to the licensee.

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