IR 05000346/1993010
| ML20045C073 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/16/1993 |
| From: | Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20045C068 | List: |
| References | |
| 50-346-93-10, NUDOCS 9306220049 | |
| Download: ML20045C073 (12) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION III
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Report No. 50-346/93010(DRP)
Docket No. 50-346 Operating License No. NPF-3
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Licensee: Toledo Edison Company Edison Plaza, 300 Madison Avenue
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Toledo, OH 43652 Facility Name: Davis-Besse Nuclear Power Station Inspection At: Oak Harbor, Ohio Inspection Conducted: April 17, 1993, through May 27, 1993 i
Inspectors:
S. Stasek R. K. Walton J. M. Shine 4 EA (> /6!93 Approved By:
b R. D. Lanksbury, Chieff Date#
Reactor Projects Section 3B
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Inspection Summary Inspection on April 17. 1993. throuah May 27. 1993
LRfp_ ort No. 50-346/93010(DRP))
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Areas Inspected: A routine safety inspection by resident inspectors of action-
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on previous inspection findings, licensee event reports, operational safety, surveillances, maintenance, and followup of events.
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Results: An executive summary follows:
Plant Operations: On April 22, 1993, operators entered Mode 4 from Mode 5 without a second makeup pump being operable as required by technical specifications (TSs)._ This occurred due -to a shift supervisor not recognizing the pump to be inoperable with its close power fuses removed.
Because of this, the inoperable equipment tracking log was not updated as required to
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include the inoperable makeup pump.
This matter was categorized as a non-cited violation (paragraph 4.b).
The subsequent approach to criticality and power ascension was conducted in a controlled manner.
On May 20, 1993,_an input failure to the integrated control system resulted in_a plant trip.
Operator intervention to prevent the trip was unsuccessful (paragraph 7.a).
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Plant recovery from the trip and.startup was also conducted in a conservative,
controlled manner. Quick operator detection ar.d response to a nonsafety-
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related component failure resulted in the prevention of a second reactor trip i-(paragraph 7.b).
A walkdown of the containment just pricr to final closecut l
9306220049 930617 PDR ADOCK 05000346 G
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revealed no substantive concerns (paragraph 4.d).
Plant housekeeping appeared f
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to be good following completion of the refuel' outage.
Radioloaical Controls: Adherence to the radiological controls program was.
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No weaknesses were identified by the inspectors.
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Maintenance / Surveillance: Maintenance and surveillance activities observed-
this period indicated the plant to be in a good state of repair.
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equipment response following a reactor trip was good with no equipment malfunctions.
However, a temporary equipment configuration change which was
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installed on check valve MS735 during testing and not removed upon' completion
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occurred late ~ in the refuel outage and resulted in a non-cited violation j
(paragraph 2.e).
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Engineerina/ Technical Support: During review of the April 22 event where a mode change was inappropriately made, it was identified that the design of the makeup pump control and indicating circuitry resulted in the " pump off" light-
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not extinguishing with the pump's close power fuses removed.
The inspector.
questioned if this design was adequate and considered the matter unresolved
(paragraph 4.b).
J Safety Assessment /Ouality Verification:
Station Review Board (SRB) and.
Company Nuclear Review Board (CNRB) meetings were observed during the
inspection period. Technical specification requirements including membership l
and quorum were met in each. case.
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DETAILS
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1.
Persons Contacted Toledo Edison Comoany
- L. F. Storz, Vice President, Nuclear and Plant Manager G. A. Gibbs, Director, Quality Assurance J. W. Rogers, Manager, Maintenance S. C. Jain, Director, Engineering
- E. M. Salowitz, Director, Planning T. J. Myers, Director, Technical Services
- J. K. Wood, Operations Manager V. J. Sodd, Manager, Independent Safety Engineering
- R. W. Schrauder, Manager, Nuclear Licensing
- G. Honma, Supervisor, Licensing
- D. R. Wuokko, Supervisor, Regulatory Affairs
- M. A. Turkal, Licensing Engineer
- R. C. Zyduck, Manager, Nuclear Engineering
- T. W Anderson, Supervisor, Maintenance Services
- M. Beier, Supervisor, Quality Assurance
- B. L. Geddes, Manager, Radiation Protection (Acting)
- B. Gallatin, Supervisor, Performance Engineering
- J. O'Neill, Maintenance Engineering D. W. Schreiner, Supervisor, Performance Engineering D. L. Eshelman, Superintendent, Shift Operations
- Denotes those personnel attending the May 27, 1993, exit meeting.
2.
Followup of Previous Inspection Findinas (92701)
a.
(Closed) Violation (346/91017-Ola(DRP1):
On September 20, 1991, the licensee commenced refilling the refueling canal and the incore instrument tank when operators noted that containment normal sump level increased at a rate of about 60 gpm. Operators found valve DH93 out of position open in lieu of closed. The root cause of this event was failure to maintain the system status file updated.
The inspectors reviewed the change to the Conduct of Operations procedure, DB-0P-00000, which was modified.to assign position responsibilities for maintaining the file and more clearly defined expectations of maintaining the file. The inspectors reviewed training records and verified that each operating crew received training on this event. The licensee implemented a program to
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improve operator-related performance after events _ attributed to operator error were identified during the sixth refueling outage.
The inspectors noted that the number of such events-have decreased. The inspectors planned to continue to monitor operator performance as part of the routine inspection program.
This violation is closed.
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b.
(Closed) Violation (346/91017-Olb(DRP)): On September 8, 1991,
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with the plant in Mode 5, the licensee experienced an inadvertent Safety Features Actuation System (SFAS) actuation while calibrating a radiation detector when a complimentary channel of SFAS was deenergized for installation of a modification.
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event was caused by inattention to detail by licensee staff coupled with a change in the outage schedule which was not reviewed for impact on plant operations.
In their post-outage critique, the licensee recognized a weakness in its review of the changing outage schedule and implemented improvements for the eighth refueling outage. The Independent Safety Engineering Group (ISEG) was responsible for evaluating schedule changes to ensure reactor safety issues were appropriately addressed during the outage. The inspectors noted that this program ensured that assumptions made during the pre-outage safety review remained valid during the outage.
The inspectors reviewed the changes to the SFAS radiation detector calibration procedures to ensure that all channels of SFAS were energized or bypassed prior to performing calibration of SFAS radiation detectors. The inspectors reviewed the licensee's final corrective actions for this event and consider this violation closed.
c.
(Closed) Violation (346/91017-02(DRP)):
On October 15, 1991, Steam Generator #2 was overfilled. The root cause of the event was operators not following procedure.
Operators elected not to perform a step in the procedure which would have vented the secondary side of the steam generator and prevented the overfill event.
Procedure DB-OP-00000, " Conduct of Operations," provided specific criteria for when procedural steps could be considered and signed off as "not applicable." However, the situation in which this allowance was applied by the operators was determined to be inappropriate for the circumstances.
The inspectors verified that the discharge from overfilling the steam generator was accounted for in the licensee's Semi-Annual Effluent Report.
The inspectors verified that operators received the training specified by the licensee in its response to the violation.
On April 15, 1993, tbr inspectors witnessed operators filling the steam generators ma verified that precautions were taken to prevent overfilling the steam generators.
This item is closed.
d.
(Closed) Vnresolved Item (346/93006-02(DRP)): Apparent inconsistencies in controlling use of overtime during the eighth refueling outage.
Further inspector review determined that although several instances were found where individuals exceeded the licensee's target internal limits, Technical Specification (TS) limits related to those persons performing or supervising
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safety-related activities were not exceeded without the required
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prior approvals. The licensee's internal control mechanism targeted a nominal 60-hour workweek during the refuel outage with any additional hours requiring appropriate management approval.
However, as a result of licensee and NRC reviews in this area, many individuals were found to be working greater than 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, but not in excess of the TS limits.
Because' regulatory limits had not been exceeded, this item was closed.
e.
(Closed) Unresolved Item (346/93006-03(DRP)): On April 10, 1993, the steam supply piping to the auxiliary feedwater turbines was filled and pressurized for testing in accordance with DB-PF-03065,
" Pressure Tests." The system was drained at the conclusion of the test and restored.
However, on April 12, 1993, performance engineers discovered check valve MS735 held open with wire and tools.
The valve was intentionally wired open to allow filling the system 2 days previously and was mistakenly not properly restored at the conclusion of the test.
Step 3.1.5 of DB-0P-03065, revision 2, required that equipment installed in the system to support testing be entered on enclosure 2.
Enclosure 2 did not indicate that MS735 was wired i
open, therefore it was not recognized at the conclusio'n of the test that the wire needed to be removed.
The condition was recognized and reported by maintenance personnel after operations l
admitted steam into the system to test the auxiliary feedwater pump turbines. With the plant in Mode 5, there were no operability requirements for the auxiliary feedwater system and testing of MS735 was not yet completed.
The licensee's corrective actions included supplemental training
for its testing staff and a review of test activities. The review-
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examined tests performed during the outage and revealed that i
temporary equipment configurations made to support testing had l
been properly made and reviewed.
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The inspectors considered this to be a violation of TS 6.8.1.c, failure to properly implement DB-0P-03065, " Pressure Tests," step 3.1.5, since a temporary modification to support testing was performed but not documented in the test. The inspectors reviewed the licensee's corrective actions and noted that safety significance was minimal since the system was not required to be operable in Mode 5 and subsequent testing of the valve would have detected the condition.
Therefore, this violation met the criteria specified in Section VII.B of the " General Statement of Policy and Procedure for NRC Enforcement Actions," (Enforcement Policy, 10 CFR Part 2, Appendix C), and will not be cited. This item is closed.
No deviations were identified in this area; however, one non-cited violation was identified.
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3.
Followup of Licensee Event Reports (92700)
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Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate i
corrective action was accomplished, and corrective action to prevent
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recurrence had been accomplished in accordance with Technical Specifications.
t a.
(Closed) LER 93001-00. Unlocked High Radiation Area Door. On February 19, 1993, a security guard performing routine checks on locked doors found that high radiation door 360 was not locked.
The licensee documented this condition on Potential Condition Adverse to Quality Report (PCAQR) 93-0057 This event was
reviewed in inspection report 50-346/93L' '(DRSS).
This LER is
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closed.
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b.
(Closed) LER 93002-00. Mode 4 Entry With Makeup Pump #1 Inoperable.
The inspectors reviewed the LER event description and the licensee's corrective actions.
Further followup will be accomplished as documented in paragraph 4.b. of this report. This
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LER is closed.
No violations or deviations were identified in this area.
4.
Operational Safety Verification (40500) (71707)
The inspectors observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during the inspection period.
The inspectors verified the operability of selected emergency systems, reviewed =tagout records, and verified tracking of limiting conditions for operation associated with affected components..
Tours of the containment, auxiliary, and turbine buildings were
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conducted to observe-plant equipment conditions including potential fire hazards, fluid leaks, and excessive vibrations, and to verify that maintenance requests had been initiated for certain pieces of equipment in need of maintenance. Walkdowns of the accessible portions of the following systems were conducted to verify operability by comparing
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system lineups with plant drawings, as-built configurations,-or present valve lineup lists; observing equipment conditions that could degrade'
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functioning, and calibrated.
-l High Pressure Safety Injection System - Trains 1 and-2
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Decay Heat Removal / Low Pressure Injection System - Train 1
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Emergency Diesel Generator No. 1
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Emergency Diesel Generator No. 2
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Component Cooling Water System (safety-related portions only)
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The inspectors, by observation and direct interview, verified that the i
physical security plan was being implemented in accordance with the
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station security plan, including badging of personnel, access control,
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security walkdowns, security response (compensatory' actions),- visitor
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control, security staff attentiveness, and operation of security equipment.
Additionally, the inspectors observed plant housekeeping, general plant -
cleanliness conditions, and verified implementation of radiation protection controls.
Specific observations and reviews included the following:
a.
The inspectors observed portions of startup activities following
completion of the eighth refuel outage. The reactor was made critical on April 28, 1993, at 11:41 p.m. (EDT). The startup was conducted in accordance with station procedures and accomplished in a conservative, controlled manner.
Except as noted in paragraph 4.b., no substantive concerns were noted.
b.
On April 22, at approximately 8:57 p.m., the unit entered Mode 4 (Hot Shutdown).
Prior to making the mode change, operators reviewed the Mode Change Checklist and the Inoperable Equipment Tracking Log to ensure all required systems and equipment were'
operable and appropriately lined up. However, although makeup
pump #1 was not included in the Inoperable Equipment Tracking Log,
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it was not operable due to its close power fuses having been
removed 2 days earlier in response to low temperature overpressurization (LTOP) procedural concerns recently identified by the licensee. The licensee documented this event in LER 93002.
The inspectors reviewed the licensee's corrective actions and
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actions to prevent recurrence documented in LER 93002 (paragraph 3.b. of this report) and determined that they were acceptable.
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Technical specification 3.0.4 specified that entry into an operational mode shall not be made unless the limiting conditions
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for operation are met without reliance on provisions contained in the action statements unless otherwise excepted. Technical specification 3.1.2.4 specified two makeup pumps shall be operable
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in Mode 4 when reactor coolant system (RCS) pressure was above 150 psig.
From about 8:57 p.m. on' April 22 until 10:32 p.m. on April 23 (when the plant reentered Mode 5 to do unrelated valve maintenance), the RCS was pressurized to about 250 psig with only one makeup pump operable.
Entry into Mode 4 with one makeup pump inoperable and RCS pressure above 150 psig was a violation of TS 3.0.4.
However, the violation was not subject to enforcement action because the licensee's efforts in identifying and correcting the violation met the criteria specified in Section VII.B of the " General Statement of Policy and Procedure for NRC Enforcement Actions," (Enforcement Policy,10 CFR Part 2, Appendix C).
A licensee review determined that the shift supervisor (SS) for the shift which removed the close power fuses did not recognize that removal of the fuses made the pump inoperable.
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makeup pumps included no automatic start functions, the SS felt
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the pump could be considered operable because it could still be
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started locally or be operated from the control room after reinstalling the fuses.
Because the pump was not recognized as i
being inoperable, an entry in the Inoperable Equipment Tracking
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Log was not made. However, the fuse removal was a shift turnover item. Additionally, a placard was placed next to the control room pump control switch indicating the close power fuses had been
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removed.
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Operator cognizance of equipment status was. inadequate at.the time of the mode change. The placard was posted on the panel for approximately 2 days before the mode change occurred and several
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operators indicated that they were aware that the fuses had been i
removed.
i From discussions with operations personnel and from review of
plant electrical drawings, it appeared that although the close power fuses had been removed, the green pump "off" light remained lit in the control room. This may have contributed to the
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operators thinking the pump was operable. The inspector questioned whether this design feature was pump specific or-generic to other pumps onsite and whether it was consistent with regulatory and/or industry standards. Additionally, NRC l
Information Notice 91-78, dated November 28, 1991, was issued i
addressing weaknesses in pump control circuits that-were identified at several facilities. The inspectors reviewed the licensee's followup of the Information Notice. At the conclusion of the inspection period,-the licensee was reviewing the. adequacy of the pump indication circuit design.
Pending resolution of the inspector's concern, this matter was considered an unresolved item
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(346/93010-01(DRP)).
c.
During the inspection period the inspectors attended Company Nuclear Review Board (CNRB) and Station Review Board (SRB)
meetings.
Technical Specification requirements including those for membership, quorum, experience levels, and use of alternates,
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Discussions were in-depth and focused on the pertinent i
issues. All members appeared to be prepared for the meeting.
d.
On April 23, 1993, the inspector performed a walkdown of containment just prior to final closecut. Overall, housekeeping was observed to be good. The licensee had conducted a general cleanup of containment and had removed or stored all items or materials with very few exceptions. As part of the general cleanup, electrical cable trays were inspected and cleaned as
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appropriate.
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No deviations were identified in this area; however, one non-cited violation was identified.
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5.
Surveillance _{61726)
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The inspectors observed safety-related surveillance testing and verified
that the testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for
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operation (LCOs) were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specification and procedure requirements and were reviewed by personnel
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other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.
The following test activities were observed and/or reviewed:
DB-MI-03002 Reactor Protection System Channel #2 Quarterly
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Calibration DB-MI-03153 Calibration of Safety Features Actuation System
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Channel #3 Radiation Detector DB-NE-03212 Zero Power Physics Testing
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DB-SC-03070 Emergency Diesel Generator #1 Monthly Run
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DB-SC-03005 Safety Features Actuation System 18-Month j
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Interchannel Logic Test
DB-SC-03077 Emergency Diesel Generator #2 184-Day Test
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DB-SP-03159 Auxiliary Feedwater Pump #2 Monthly Jog Test
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No violations or deviations were identified in this area.
6.
Maintenance (62703)
Station maintenance activities of safety-related systems and components were observed'and/or reviewed during the inspection period to ensure
that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance with Technical Specifications.
The following items were considered during this review:
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conditions for operation (LCO) were met while components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were
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performed prior to returning components or systems to service, quality
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control records were maintained, activities were accomplished by qualified personnel, parts and materials used.were properly certified, radiological controls were implemented, and fire prevention controls-
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were implemented.
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Maintenance work orders (MW0s) were reviewed to determine the status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which may affect system performance.
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The following maintenance activities were observed and/or reviewed:
MWO 7-93-0264-01 Remove, Disassemble and Repair SW1429
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MWO 3-93-0722-06 Preventive Maintenance on #2 EDG
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MWO 2-91-0046-47 -Rewire Node 3, Fire Detection System
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No violations or deviations were identified in this area.
7.
Followun of Events (937011 During the inspection period, the licensee experienced two events, one of which required prompt notification of the NRC pursuant to 10 CFR
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50.72. The inspectors pursued the events onsite with the licensee and/or other NRC officials.
In each case, the inspectors verified that i
the notification was correct and timely, if appropriate, that the
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licensee was taking prompt and appropriate actions, that activities were
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conducted within regulatory requirements, and that corrective actions
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would prevent recurrence.
The specific events were as follows.
a.
On May 20, 1993, at about 11:35 a.m. (EDT), with the plant operating at 100 percent power, a malfunction.in the T.. input to the plant integrated control system (ICS) occurred. The T,, input to ICS failed to its midposition (about 570 F) generating a-
control rod-withdraw demand during which ICS attempted to restore T..
to 582 F.
This resulted in reactor power increasing to about i
103 percent at which point the ICS' Reactor Demand High Limiter prevented further rod withdrawal.
Reactor coolant system pressure j
increased until pressurizer spray was initiated. Operators stabilized the plant at about 103 percent power (TS trip setpoint
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was less than or equal to 104.94 percent of rated power) and normal operating pressure. About 3 minutes after the initial
failure, control room operators manually reduced the ICS demanded
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power which reduced feedwater flow but did not have the
anticipated effect of reducing the control rod demand signal since the signal was modified by the errant T.. input. With reduced j
feedwater. flow and no change in' control rod position, reactor J
pressure increased. At 11:38 a.m., the plant tripped on a high j
reactor coolant system pressure signal from the Reactor Protection i
i System (RPS).
i Prior to the trip, the safety parameter display system (SPDS) was j
unavailable, but did not interfere with the operators' ability to
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collect plant data.
Post-trip response was normal.
The licensee determined that a loose power supply fuse in the T. input to ICS
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j caused the T.., signal to fail to midscale.
The licensee replaced the faulted fuse and cap, then inspected other related fuse assemblies to determined if a similar condition existed elsewhere.
All fuse assemblies checked were determined to have no loss of
continuity.
However, some fuse caps and fuses were replaced as a result of the visual inspection. ' The licensee initiated ~an
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investigation into the details of both plant and operator response J
to thr, event.
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The licensee documented the reactor trip on PCAQR 93-0311 to evaluate the long term corrective actions to prevent recurrence of
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this event. The licensee completed its post-trip review of the
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event and determined that the plant could be restarted. The plant was made critical on May 21, 1993, at 7:14 a.m., and full power was achieved at 10:31 p.m.
The plant recovery and subsequent
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startup was performed in a conservative, controlled manner without problems.
The inspector noted that control room operators quickly detected I
the symptoms of the ICS problem, but were unable to immediately
determine if the T.. input to ICS was a faulted or actual condition.
Because of this, the operator response to the condition unexpectedly produced a high RCS pressure condition which ultimately resulted in a reactor trip.
Plant equipment and systems responded per design and within analyzed conditions.
This was indicative of good overall material condition at the facility.
The failure of the fuse cap in the T,.
input to the ICS was similar to the failure of an identical fuse cap in the RPS, i
(documented in PCAQR 92-0090).
The licensee's corrective actions
did not address similar fuse caps in other applications.
The licensee did not consider a single component failure as
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necessitating review of all similar components unless there was_ a i
known generic deficiency. At the conclusion of the inspection period, the licensee was in the process of determining if any other systems use this model fuse assembly and whether more extensive corrective actions were appropriate, l
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b.
On May 24,1993, at.12:21 p.m., control room operators detected a decreasing level in the Turbine Plant Cooling Water (TPCW) high level tank.
The TPCW system supplied gravity-driven cooling to nonsafety-related components in the turbine building from the high-level tank. A decreasing level in the TPCW high level tank was a precursor to a loss of TPCW cooling and necessitated operators to reduce power to prevent overheating TPCW cooled components.
Pl ant power was reduced from 100 percent to about 76 percent before operators detected and bypassed a faulty TPCW high level tank level control valve (CW620). Tubing between the air-operated i
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level control valve and controller had broken due to an apparent vibration induced fatigue failure. The licensee refilled the TPCW high level tank using a bypass valve and returned the plant' to 100 percent power at 2:09 p.m.
The inspectors noted that quick detection of the faulted valve and initiation of compensatory actions by the operators prevented a
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No violations or deviations were identified in this area.
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8.
Unresolved Items
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An unresolved item is a matter requiring more information in order to ascertain whether it is an acceptable item, a violation, or a deviation.
During this inspection, an unresolved item was identified (paragraph 4.b).
9.
Violations For Which A " Notice of Violation" Will Not Be Issued The NRC uses the Notice of Violation to formally document failure to
meet a legally binding requirement. However, because the NRC wants to
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encourage and support licensee initiatives for self-identification and
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correction of problems, the NRC will not issue a Notice of Violation if
the requirements set forth in Section VII.B of the " General Statement of
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Policy and Procedure for NRC Enforcement Actions," (Enforcement Policy,
10 CFR Part 2, Appendix C) are met. Violations of regulatory l
requirements identified during the inspection for which a Notice of
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Violation will not be issued are discussed in paragraphs 2.e and 4.b.
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10.
Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1) throughout the inspection period and at the' conclusion of
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the inspection on May 27, 1993, and summarized the scope and findings of
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the inspection activities. The licensee acknowledged the findings.
l After discussions with the licensee, the inspectors determined there was no proprietary information contained in this inspection report.
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