ML20197J367

Insp Rept 50-298/86-12 on 860318-21.Violation Noted: Switchgear Cabinets for Emergency Diesel Generators Not Anchored by Welding or Bolting

ML20197J367
Person / Time
Site:	Cooper
Issue date:	05/07/1986
From:	Dubois D, Jaudon J, Mcneill W, Plettner E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20197J360	List: ML20197J356 ML20197J363 ML20197J367
References
50-298-86-12, NUDOCS 8605200009
Download: ML20197J367 (8)

See also: IR 05000298/1986012

Text

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. APPENDIX U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report: 50-298/86-12 License: DPR-46 Docket: 50-298 Licensee: Nebraska Public Power District (NPPD) P. O. Box 499 Columbus, NE 68601 Facility Name: Cooper Nuclear Station (CNS) Inspection At: Cooper Nuclear Station, Nemaha County, Nebraska Inspection Conducted: March 18-21, 1986 Inspectors: I. 4. [4df.o 3 /3///4 E. A. Plettner, Resident Inspector, (RI) Date W%k 3 / 91/ 2 6 D. L. DuBois, Senior Resident Inspector, (SRI) Dats U/7Yt N b W. M. McNeill, Project Erfgiheer, Project Da te' / Section A, Reactor Projects Bran Approved: . &/ [J4 C7M J. P Jau on,~ et, Projdct Section A, Date / R cto Proj t Branch 8605200009 860513 DR ADOCK 0 20

' . 2 Inspection Summary Inspection conducted March 18-21, 1986 (Report 50-298/86-12) Areas Inspected: Routine, unannounced inspection of onsite followup of written reports of nonroutine events. The two events concerned the lock of seismic support of emergency power switchgear cabinets and the removal from service of components in one safety system train while the redundant safety train dedicated emergency clesel generator was inoperable. Results: Within the areas inspected, one violation was identified (failure to meet seismic requirements). .

. 3 DETAILS 1. Persons Contacted

• G. R. Horn, Division Manager of Nuclear Operations
• E. M. Mace, Plant Engineering Supervisor
• J. M. Meacham, Technical Manager
• V. L. Wolstenholm, Quality Assurance Manager
• Denotes presence at exit interview.

2. Lack of Seismic Support of Emergency Power Switchgear Cabinets On March 18, 1986, the licensee informed the onsite NRC resident inspectors that two switchgear cabinets associated with the No. 2 emergency diesel generator (DG2) did not appear to be secured to the respective floor mounting plates as required to meet seismic criteria. The affected cabinets included DG2 output breaker (EG2) cabinet and DG2 auxiliary control cabinet. This deficiency was discovered on March 18, 1986, by licensee personnel during their performance of Special Procedures (SP) 86-002, "As-Building of EG2 Cabinet and DG2 Cabinet in DG2 Room", dated February 13, 1986, and SP 86-002, Amendment 1, " Structural As-Building of DG2 Control Panel in DG2 Room", dated March 10, 1986. Immediate licensee actions included the following: The underside of the affected cabinets were reportedly probed with a . steel rule which verified the lack of welds or bolts for securing the cabinets to the floor mounting plates. Visual inspections were performed of other electrical cabinets located . in the DG2 room. These inspections verified that the other electrical cabinets were mounted by bolting and/or welding. The EG2 cabinet and DG2 auxiliary control cabinet were welded to their . respective floor mounting plates using a combination of plug and fillet type welds. This occured on the same day as the lack of bolts or welds was discovered. Upon completion of the above, the licensee returned DG2 to service and tagged out DG1 in order to perform an inspection of how its associated cabinets were mounted. The licensee discovered that the DG1 output breaker (EG1) cabinet and DG1 auxiliary control cabinet were alsc not secured to the respective floor mounting plates. The licensee weldtd the affected cabinets to their respective floor mounting plates during the evening of March 18,1986. Upon completion, DG1 was returned to an operable status. The licensee performed followup actions during this inspection period which included: Preliminary seismic analysis of the corrective welds. . L

,. . . -- .-- . . . - , ~ ..- - - _ . _- __- - l - , 4 Additional inspections of other electrical switchgear cabinets, . including all motor control centers located on elevation 903 ft. of the reactor building, a sampling of five 4160 V AC breaker cabinets located in critical switchgears IF and 1G, 480V AC critical load centers F and G, and 4160V AC noncritical switchgears. No additional panel mounting discrepancies were discovered. Nonconformance Report (NCR) 005032 was written On March 18, 1986, to . document the licensee findings in this area. The Station Operations Review Committee (SORC) met on March 18 and 19, . 1986, to discuss their inspection findings and immediate corrective actions. (ref: 50RC Meetings #469 and #470 minutes). Upon notification by the licensee of the above identified problems, the NRC inspectors performed the following actions during this inspection period: Reviewed a General Electric (GE) instruction manual for metal clad . l switchgear equipment (GEK-32770) and verified that the recommended method for installing or anchoring the cabinets was by welding with an 4 alternate method by bolting. , " Noted that not all applicable pages GE drawing 0673D0515 referenced in . the instruction manual were available at the time of the inspection. It is assumed that this drawing illustrates vendors recommendations l for anchor points, weld size, and weld length. - { Reviewed Maintenance Work Requests (MWRs) 1227 and 1228 which documented . corrective welding. The licensee welded four (two only in EG1) plug . + type welds inside the front of the cabinets and applied fillet type - welds across the full length of the backside of the cabinets. In addition, intermittent fillet welds were applied to the front of the cabinets.. Examined the final, as-welded condition of the fillet welds. . a. ' Reviewed the special inspection procedures SP 86-002 and SP 86-002 . Amendment 1 for technical adequacy. . Reviewed surveillance tests to verify that Technical Specification . required tests to demonstrate operability were performed prior to, ' - during, and following the licensee's special inspections. Verified that affected equipment was properly safety-tagged. . ' Visually inspected additional electrical cabinets to determine if the . method of mounting them was in accordance with drawings. - Verified that maintenance activities were performed and documented in . i accordance with licensee administrative procedures. i < . - - , ---,m. ,. .m ,, , , , - . . , , - , . - - - - - - - - - . - - - , , _ - - - - - - , . , _ , . , - - - _ , , .

" . 5 Further licensee actions remaining to be completed are: The completion of seismic evaluations of the welds made as corrective . action to NCR 005032. Visual inspection of additional safety-related electrical cabinets for . support inadequacies. A review of construction records to establish the cause of the mounting . anomalies and to determine if other anomalies exist. Final disposition of NCR 005032. . Write a Licensee Event Report (LER) and submit to the NRC by April 17, . 1986. The CNS Technical Specification, section 5.6, establishes that Class I structures and equipment are to meet seismic design requirements. The Updated Safety Analysis Report in regard to " Emergency Power System" , requires that equipment be capable of withstanding seismic conditions. The failure to identify this construction deficiency to meet seismic requirements was identified as an apparent violation (298/8612-01). 3. Lack of Safety-related Component Redundancy On March 18, 1986, the safety system division II emergency Diesel Generatnr (DG2) was rendered inoperable in order to perform special inspections SP 86-002 and SP 86-002, Revision 1. DG2 is the source of emergency power for "C" and "D" Reactor Equipment Cooling (REC) Pumps, whereas DG1 supplies emergency power to Division I REC Pumps "A" and "B". During the period that DG2 was inoperable, licensee operations personnel rendered "A" and "B" Pumps inoperable, one at a time, to implement a design change to the pump casing vents. CNS Technical Specification section 3.12.B.2, " REC System", states, "From and after the date that any component in one loop becomes inoperable continued reactor operation is permissible during the succeeding thirty days provided that during such thirty days all the components of the other loop and the active components of the engineered safeguards compartment cooling systems, the diesel generator associated with the operable loop are operable." Technical Specification section 3.12.8.4 states that, if the preceding requirement can not be met, the reactor will be shut down and placed in the Cold Shutdown condition within 24 hours. "A" REC pump was inoperable for 2 hours and 25 minutes and "B" REC pump was inoperable for 45 minutes as indicated by clearance orders 86-229 and 86-231 respectively. During this period the emergency power supply to "C" and "D" REC pumps was not available because DG2 was tagged out of service. _-

l . 6 Licensee personnel did not recognize that the Technical Specification quoted above meant that neither "A" nor "B" pump should be taken out of service with DG2 out of service. However, during the time the "B" REC pump was inoperable, the operations supervisor identified the abnormal situation and took imediate steps to have that pump returned to an operable condition. Followup licensee corrective actions included: Notification of the event to the SRI and the NRC operations duty . officer. An immediate issuance of Special Order 86-02, dated March 18, 1986, . which imposed more stringent requirements to operating crews prior to removing Technical Specification related equipment from service. A copy of Special Order 86-02 was routed to the maintenance department. . NCR 006046 was written on March 18, 1986, to document the occurrence . and to recommend its disposition. The SRI performed the following actions concerning this event: Verified completion of the imediate and followup actions performed by . the licensee which are listed above. Ensured licensee adherence to other related Technical Specification . requirements. Interviewed two senior reactor operators and two reactor operators . from the affected operations crew. The SRI determined from those interviews that not one of the operators recognized the Technical Specification requirements stated above at the time of the event. Interviewed the operations and maintenance supervisors concerning the . conditions leading up to the event. The length of time that the "A" and "B" REC pumps were out of service was within that allowed by the Technical Specifications. Therefore, this is not considered to be a violation. The NRC inspector concluded that this was an indication both of a breakdown in management control and an evident l weakness in operator knowlea9a 4. Exit Meeting The NRC inspectors conducted an exit meeting on March 21, 1986, with the licensee personnel denoted in paragraph 1. At this meeting, the purpose and the findings of the inspection were summarized.

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