IR 05000346/1986025

Insp Rept 50-346/86-25 on 860819-22 & 0904.No Violation or Deviation Noted.Major Areas Inspected:Inservice Insp Activities & Resolution of Previous Insp Findings

ML20206R682
Person / Time
Site:	Davis Besse
Issue date:	09/15/1986
From:	Danielson D, Jeffrey Jacobson, David Jones NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206R677	List: IR 05000346/1986025 ML20206R674
References
50-346-86-25, NUDOCS 8609190139
Download: ML20206R682 (6)

Text

_ _ . _ _ . _ _ _ . _. ._ _ _ _ -._ .. _ -

%

.

- -

,-

,

~

- U.S. NUCLEAR REGULATORY COMISSION =

REGION III

- Report No. 50-346/86025(DRS)

' Docket No. 50-346- License No. NPF-3 Licensee: Toledo Edison Compan Edison Plaza, 300 Madison Avenue Toledo, OH. 43652 4

'

i-Facility Name: Dr/is-Besse Nuclear Power Station, Unit 1

- Inspection At: Davis-Besse Site, Oak Harbor, OH ';

( - Inspection Conduct d: August 19-22 an'd September 4, 1986

"

c~

!

Inspector: . s 64 T- /f-f6

. Date

.

W

! . . aco son 7-/P f6 W

Approved By: D. H. Danielson, Chief _

4 IT 8%

'

Materials and Processes Section Date i

,

Inspection Summary *

'

' Inspection on August 19-22 and September 4, 1986 (Report No. 50-346/86025(DRS))-

Areas Inspected: Routine, unannounced inspection of inservice inspection

'

(ISI) activities and of the resolution of previous inspection findings. This report addresses the-following inspection modules: 73051, 73052, 73753, 73755, 9270 Results: No violations or deviations were identified.

J I D

,

I B609190139 609158 PDR

I G ADOCK 05000346 i B PDR .

,

t '

,- ---~e , , - , , n--, , ~ , . , - .n+ ,e r--.n,-. nrm n-,-w.w-.--,- ,.n-,.,- .,4

. ,mngoc ,.-av-,n-g g._,mycrn----g,w-,-,,wn--g_,,.- < , ,

-.

'

<

~*

.

DETAILS

, Persons Contacted Toledo Edison Company (TEco)

*L. Ramsett,'QA Director

• C. Daft, Operations QC Manager

• M.' Shepherd, Code Inspection Supervisor T. Thompson, Outage Supervisor Babcock and Wilcox-Company (B&W)

M. Storey, Eddy Current Task Leader -

C. Webber, Senior Analyst K. Matthews, S/G Tube Plugging Task Leade Hartford Steam Boiler Inspection and-Insurance Company B. Hogstrom, Authorized Nuclear Inservice Inspector U.S. Nuclear Regulatory Commission (NRC)

~* Byron,. Senior Resident Inspector D. Kosloff, Resident Inspector The inspector.also contacted and interviewed other licensee and contractor. employee * Denotes those'present at the exit intervie . . Licensee Action on Previous Inspection Findings (Closed) Open Item (346/85-15-01): "CRDM Spring Failure Hydrogen Damage Report" The NRC inspector reviewed the licensee's report r

'

titled " Control Rod Drive Mechanism Leaf Spring Failure Mode Validation." This report examines the leaf spring material, fabrication, service and failur ,

An engineering evaluation was performed to determine the most probable cause of the leaf spring failures. This evaluation consisted of a literature survey to determine corrosion and mechanical properties of the leaf spring material, a failure analysis, and a dimensional and structural evaluation of the CRDM components as they relate to the leaf spring.

i From the engineering evaluation, it was determined that the most

.

probable cause of the leaf spring failure was due to mechanically

'

induced bending stress. It was determined from the literature that l

'

the 17-7PH (TH1050 condition) material used to fabricate the leaf spring would not be sJsceptible to corrosion mechanisms in a

,

!

}- 2 i

i

. _ , . . _ . . . _

_ . . _ , _ . . _ . , . _ . _ , _ . _ _ . _ . . , _ , . _ . . . , _ . . , _ . _ _ . - _ . _ . _ - . , , , . . , . _ . _ . . _ _ - . . ,

.

.

..

simulated reactor environment' stressed below yield. Failure analysis examinations _ performed on failed leaf springs from Davis-Besse.showed that the 17-7PH material gave. expected micro-

. structure, chemistry, and hardness results. However, several

'

scratched areasion the leaf springs relating to deformed or chipped-areas on the torque tube cap were noted. In addition, the servic leaf spring failure was duplicated in the laboratory using a three point bend test in 200*F water. Dimensional:and structural evaluation'of the CRDM components showed that if the locking pin is not. properly engaged in the slot, the leaf spring failure was duplicated in the laboratory using a three point bend test in 200*F wate Dimensional and structural evaluation of the CRDM components-showed that_if the locking pin is not ptoperly engaged in the slot, the leaf spring can interfere with the torque tube cap in the driv This~ interference would be expected to cause failure of the high-strength notch sensitive. leaf spring materia It was' determined that failures of this type can be prevented if the leaf spring remains in the locked position. It was recommended by the report that all leaf springs be inspected once to determine they

'are in the latched position and free of fracture. Thereafter, only

. verification that the leaf springs have been relatched would be necessar To assure that the springs were currently in an unbroken state, all of the CRDM leadscrews except the APSRs were raised and inspected to verify that the spring was present, was correctly seated within its locking slot, and that the spring retained its " springiness."

This~ inspection revealed that all but three of the remaining locking

. pins were in place. Of the five springs which were found unlocked at the Davis-Besse site, four were inserted into their locked position, and one was removed from the CRDM with a through crack found at the rivet hole nearest the tip of the spring. This crack ran.from one outside edge to the rivet hol Since that time, all of the leadscrew nut--locking springs which were found in the extended position were removed from their drive to assure that no potentially damaged springs were placed back into service.- Dye penetrant inspection of a number of these have not shown any cracks. Additionally, six other good leadscrew nuts were removed to allow visual. inspection of the springs. No cracks were detected, (Closed) Open Item (346/84-28-06): " Maintenance Procedure Needed for RCP Cover Removal" the NRC inspector reviewed the licensee's maintenance procedure, MP 1700.83, " Reactor Coolant Pump Disassembly, Inspection, Repair and Reassembly," Revision 2. The procedure addressed the criteria of IE Bulletin 82-0 (Closed) Open Item (346/84-28-07): "MT/PT Inspection Practices on RCS Threaded Fasteners" the NRC inspector reviewed the licensee's maintenance procedure, MP 1700.83, " Reactor Coolant Pump Disassembly,

. . .

.

Inspection, Repair and Reassembly," Revision 2. The procedure addressed visual and magnetic particle inspection requirements for threaded fastener . Steam Generator Eddy Current Examination General Information Toledo Edison Company contracted Babcock =and Wilcox to perform the steam generator eddy current examination in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, 1977 Edition through Summer 1978 Addend Eddy current examination was performed on steam generators (SG) A and B and the following tubes were plugged:

SG A Row Tube- Row Tube 68 67 82 66 70 69 82 69 70 70 83 67 71 63 84 63 71 71 85 64 73 52 85 65 75 65 85 70 78 65 87 66 82 64 -88 67 SG B Row Tube 76 62 Plan / Procedure Review The inspector reviewed selected portions of the following plan / procedures:

• B&W, Technical Procedure for the Evaluation of Eddy Current Data of Nuclear Grade Steam Generator Tubing, ISI 460, Revision 1 * B&W, OTSG EC Inspection Plan Davis-Besse 1, A-0TSA, August 1986 Inspection, 51-1164117-0 * B&W OTSA EC Inspection Plan Davis-Besse 1, B-ITSA, August 1986 Inspection, 51-1164118-0 .-

.

• B&W, Technical Procedure for the Multifrequency Eddy Current Examination of OTSG Tubing in 177 Steam Generators Using the MIZ-18, ISI-418, Revision * B&W, Eddy Current Examination of Tubing by the Absolute Multicoil (8x1) Technique, ISI-425, Revision 1 * B&W, Technical Procedure for.the Evaluation of Eddy Current Data for Debris, Sludge and Secondary Side Loose Parts i'n Steam Generators, ISI-462, Revision * B&W, Technical Procedure for the Evaluation of Eddy Current Data Generated From the Multielement Probe, ISI-463, Revision * B&W, Technical Procedure for the Evaluation of Eddy Current Data of Nuclear Grade Steam Generator Tubing for Intergranular Attack, ISI-467, Revision * B&W, Eddy Current Procedure for Determining the Clearance Between OTSC Tubes and the Internal Auxiliary Feedwater lleader, ISI-413, Revision * B&W, Insertion, Calibration, Operation, and Removal of Eddy Current OTSG Tube Examination Equipment in the Upper Head, 151-406, Revision s No violations or deviations were identifie c. Material and Equipment Certification The inspector reviewed the certification documents, relative to the following items:

• Eddy current equipment No violations or deviations were identified, d. NDE Personnel Certifications and Observation of Work Activities The inspector reviewed NDE Personnel Certification in accordance with SNT-TC-1 The inspector also observed the work had discussions with personnel during the following examination:

• Eddy current examination of steam generator A, Row 92 - Tube 99, Row 91 - Tube 94, Row 84 - Tube 82, Row 89 - Tube 91 and Row 103 - Tube 105 in accordance with B&W Procedure ISI-418, Revision 2, " Technical Procedure for the Multifrequency Eddy Current Examination of OTSA Tubing in 177 Steam Generators using the MIZ-18."

.

.

No violations or deviations were identifie Data Review The NRC inspector reviewed report documentation relative to the following:

• Eddy current examinations No violations or deviations were identifie . Exit Interview The inspector met with site representatives (denoted in Persons Contacted paragraph) at the conclusion of the inspection. The inspector summarized the scope and findings of the inspection noted in this report. The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection. The licensee did not identify any such documents / processes as proprietar ,

6