IR 05000346/1993004
| ML20035C738 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 04/02/1993 |
| From: | Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20035C733 | List: |
| References | |
| 50-346-93-04, 50-346-93-4, NUDOCS 9304090049 | |
| Download: ML20035C738 (11) | |
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U. S. NUCLEAR REGULATORY COMMISSIO!4
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REGION Ill Report No. 50-346/93004(DRP)
Docket No. 50-346 Operating License No. NPF-3 Licensee: Toledo Edison Company Edison Plaza, 300 Madison Avenue Toledo, OH 43652 Facility Name: Davis-Besse Nuclear Power Station Inspection At: Oak Harbor, Ohio inspection Conducted: January 26, 1993, through March 15, 1993 Inspectors:
S. Stasek R. K. Walton-Approved By:
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4f2M R. D. lanksbury, Chief Date Reactor Projects Section 3B Inspection Summary Inspection on January 26. 1993, through March 15. 1993 (Report No. 50-346/93004(DRP))
Areas Inspected: A routine safety inspection by resident inspectors of action on previous inspection findings, operational safety, preparations for
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refueling, surveillances, and maintenance.
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Results: An executive summary follows:
Plant Operations: Overall, performance of the operating crews was again good'
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The shutdown and cooldown of the unit to begin the.
eighth refueling outage was conducted in a controlled, conservative manner.
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In general, adherence to administrative controls was also good.
However, two j
incidents occurred during -activities assstiated with new fuel receipt ~
l operations in the spent fuel pool (SFP) h ea (paragraph 4).
The first
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involved a SFP bridge inadvertent " bump" of the west transfer mechanism winch
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platform during a normal bridge traverse.
The second involved a new fuel l
assembly inadvertently " catching" on the upper frame guide.of the new fuel
elevator as it was being lowered into the elevator using the overhead crane.
This caused the assembly to swing laterally approximately 6 inches into the.
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west transfer trechanism's motor housing.
Both incidents involved aspects of
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operator error and inadequate attention-to-detail.
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9304090049 930402 PDR ADOCK 05000346
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Radioloaical Controls:
In general, adherence to radiation protection program
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requirements-was good this period. However, the containment emergency airlock
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access door, which was required to be locked as a high radiation area door, was I
found not properly secured (paragraph 3.e). Further followup of this matter has i
been conducted by the Division. of Radiation Safety and Safeguards (reference
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Inspection Report 50-346/93007(DRSS)).
Maintenance / Surveillance:
Overall, maintenance and surveillance activities observed this period appeared to be conducted in accordance with all applicable
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licensee. requirements.
However, some weaknesses were initially noted ' and subsequently corrected involving erection of scaffolding in safety-re. lated areas of the plant (paragraph 3.b). In addition, a question was raised relating to the NRC endorsed ventilation testing methodology used to meet the technical
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specification surveillance requirements.
Specifically, the American National
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Standards Institute (ANSI) Standard referenced in the surveillance requirements through a regulatory guide allows some adjustments to be made prior to conducting
certain tests.
This appeared to be inconsistent with NRC restrictions on
preconditioning equipment prior to test performance (paragraph 5.b). This matter will be tracked as an inspection followup item.
- r Enaineerina/ Technical Support:
Engineering support prior to and during the refueling outage appeared to be timely and, overall, responsive to the other site z
department's needs.
The inspectors noted no substantive concerns during this inspection period. A review of how the licensee had determined low temperature overpressurization (LTOP) limits and setpoints was conducted with no outstanding issues identified (paragraph 3.c).
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1.
Persons Contacted a.
Toledo Edison Company D. C. Shelton, Vice President, Nuclear
- G. ~A. Gibbs, Director, Quality Assurance
- L. F.. Storz, Plant Manager
- S. C. Jain, Director, Engineering
- E. M. Salowitz, Director, Planning J. K. Wood, Operati.ons Manager
- J. R. - Polyak, Manager, Radiological Protection
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- G. M. Grime, Manager, Industrial Security
- R. W. Schrauder, Manager, Nuclear Licensing.
- G. Honma, Supervisor, Licensing-
- N. K. Peterson, Engineer, Licensing
- D. R. Wuokko, Supervisor, Regulatory Affairs
- R. C. Zyduck, Manager,. Nuclear Engineering
- T. J. Myers, Director, Technical Services
- D. W. Schreiner, Supervisor, Performance Engineering.
- D. L. Haiman,-Manager, Engineering Assurance and Services
- W. C. Rowles, Independent Safety'and Nuclear
- D. L. Eshelman, Superintendent, Shift Operations
- N. L. Bonner, Manager, Design Engineering
- C..A. Hengge,. Supervisor, Systems Engineering
- F. W. Zurvalec, Systems Engineering
- A. L. McAllister, Systems Engineering
- T. S. Swim, Supervisor, Design Engineering b.
- S. Stasek, Senior Resident Inspector
- R. K. Walton, Resident Inspector
- Denotes those personnel attending the March 15, 1993, exit meeting.
2.
Followuo_of Previous Inspection Findinas (92701)
a.
(Cloted) Violation (346/92002-02(DRP)):
Both hydrogen monitors made inoperable without control room operators knowledge.
In response to this event,.the licensee _~added labels and improved locks on the access doors to the cabinets to reduce the potential-of inadvertent v eration of the monitors and their power supply breakers. The licensee also initiated a program to evaluate and remove unnacessary, redundant, or unuseful computer alarms displayed on the control room alarm ' monitor. ' The licensee currently plans to remove about 2300 of these computer alarm points from the monitor with the program expected to be completed in 1994. -The computer alarms removed from the monitor.will
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continue to be displayed on the alarm printer located in the control room. Operators are required to check the printout twice per shift.
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The inspector reviewed the computer alarm monitor and found that
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the number of alarms on the monitor had been substantially reduced (from four screens in length to less than one screen).
In general, operators indicated that the reduced number of alarms on the monitor were an improvement and made the monitor a more useful tool. This item is closed.
b.
(Closed) Open Item (346/92003-01(ORP)):
Failure to maintain the shift manager's log in accordance with administrative requirements. The licensee subsequently issued a change to the administrative procedures which eliminated the reactor operator log and shift manager log.
The one remaining log (i.e., the unit log), was to be maintained by the assistant shift supervisor and would include information that had formally been entered in the previous logs. The inspectors have reviewed the unit log on an ongoing basis and have noted an improvement in logkeeping practices.
It was noted that all members of the control room operating crew still retain the capability to make entries to the unit log. This item is closed.
c.
(Closed) Open Item (346/92019-01(DRP)): Revision to
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administrative controls to provide more specific guidelines to i
operators regarding procedural adherence. The licensee subsequently revised administrative procedure DB-0P-00000,
" Conduct of Operations," to include more specific guidelines on procedural adherence requirements. This item is closed.
No violations or deviations were identified in this area.
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3.
Operational Safety Verification (71707) (40500)
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.
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Tours of the containment, auxiliary, and turbine buildings were conducted to observe plant equipment conditions including potential fire h. 'ards, 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 configuration, or present valve lineup lists; observing equipment conditions that could degrade i
performance; and verifying that instrumentation was properly valved, functioning, and calibrated.
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Service Water System (safety-related portions)
The inspectors, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the 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
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equipment.
Additionally, the inspectors observed plant housekeeping, general plant
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cleanliness conditions, and verified implementation of radiation protection controls.
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.I Specific observations and reviews included the following:
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a.
On February 28, 1993, at approximately 11:30 p.m., the licensee commenced a reactor shutdown to begin the eighth refueling outage.
.i The inspector observed portions of the shutdown and cooldown of
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the reactor with no substantive concerns noted. Overall, control room activities were well controlled with the shutdown and cooldown evolution being conducted in a conservative,- orderly
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manner.
b.
On February 9,1993, during a routine plant tour, the inspector noted instances where scaffolding had been erected in close proximity to, or in contact with, safety-related equipment.
Specifically, one scaffold in mechanical penetration room #4 was
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found with installed clamps wedged against cable tray BLGFIBN.
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Also, in mechanical penetration room #2, ~ a scaffold was found in contact with a safety-related instrument line.
In both cases, the scaffolding was firmly anchored in place, and would not adversely affect these plant components.
Further review by the licensee subsequently determined-that it was l
unclear whether a review of the scaffolding configuration had been
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additional clearance.
In addition, the scaffolding installation i
log was revised to assure proper reviews would be documented as
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such in the future.
Subseq' ently, a quality assurance
u surveillance (SR-93-MAINT-01) was conducted which found the
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program was adequately implemented. The inspector then conducted.
additional walkdowns with no furthe~r weaknesses noted.
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i c.
The inspectors responded to a regional request to evaluate the
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licensee's methodology to prevent a low temperature _
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overpressurization (LTOP) event. 'The inspectors reviewed the licensee's previous HRC submittals and associated technical specification requirements. The inspectors found the Babcock and Wilcox Company (B&W) analysis, technical specifications, in-plant '
equipment, and administrative controls were consistent.
Specifically, pressure relief valve DH-4849, located in the _ decay
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heat suction piping, was utilized to prevent primary system pressure from exceeding 330 psig. The setpoint'was conservative to the B&W analysis. Control room indications of' system pressure I
and temperature to monitor for LTOP limits were selected
appropriately.
In addition, the licensee determined that under i
low pressure / temperature conditions, the makeup system would need
to be disabled to assure that it would not further contribute to
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an overpressurization event. The inspectors had no concerns with the licensee's engineering evaluations nor the establishment of
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operating limitations in the form of curves and relief valve
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setpoint determination.
d.
The inspectors responded to a regional request to determine I
whether an ethylene glycol solution was used in any of the diesel generator (DG) jacket water cooling systems onsite. The concern
was that since ethylene glycol provided less cooling capability
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than water, the DG full load rating may have to be reduced from i
that originally assumed. The licensee indicated that only the
statinn blackout diesel generator (SBODG) used ethylene glycol and
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engineering personnel contacted the DG vendor to ensure that the engine's design full load rating had accounted for the use. On
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February 2,1993, the licensee reported that the SBODG was a newer -
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model and that the use of ethylene glycol did not affect the
current output rating of the engine.
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e.
At 4:03 p.m. on February 19, 1993, a security guard performing
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routine checks on locked doors, found that door number 360, access i
door to the containment emergency escape hatch, was not locked.
The door was required to be locked since it accesses a high
radiation area (i.e., containment). The door initially was believed to have been locked since it did not move when pulled,
but upon further investigation, the dead bolt was not visibly in
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contact with the latch indicating that the locking mechanism was
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not engaged. The door was then opened to verify that the lock was not engaged, then locked closed.
Since this door was alarmed, a
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review of the computer records was done that indicated the door i
had not been operated since the evening of February 17, 1993.
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On February 17, at 7:22 p.m., after the successful completion of
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the emergency escape hatch local leak rate test, door 360 was
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closed by security and two radiological controls technicians, all i
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of which believed the door automatically locked when it was closed. They pulled on the door to confirm that the door had
locked, then left the area. Hourly security checks of the door
did not detect that the door was not locked until the afternoon of
February 19, 1993.
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i The licensee documented this event in a Potential Condition
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Adverse to Quality Repurt (PCAQR 93-0057) and will evaluate j
corrective actions and reportability.
Further NRC followup of i
this matter has been conducted by the Division of Radiation Safety and Safeguards (reference Inspection Report 50-346/93007(DRSS)).
f No violations or deviations were identified in this area.
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Preparations for Refuelina (60705) (60710)
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During the inspection period, the inspectors observed selected portions
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of receipt inspection, handling, and storage of new fuel and control
components. The shipments arrived during January and February 1993 in
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support of the upcoming refueling outage that subsequently started on
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March 1, 1993.
The inspectors observed the licensee's accountability,
radiological controls, quality control, and housekeeping programs during i
receipt operations.
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Overall, the activities observed were adequately conducted in a o
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controlled manner. However, on February 2, 1993, during receipt inspection activities, the licensee noted that the identification serial numbers on two of the new fuel bundles housed in one shipping container
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did not match those on the shipping papers, including DOE /NRC Form 741, i
" Nuclear Material Transaction Report (NMTR)." Further review ensued i
with the determination made that the fuel supplier (B&W) had
inadvertently switched the papers associated with the container with those from another container scheduled to arrive at the site later in the week.
In both cases, the overall amount of-special nuclear material shipped was the same because the fuel composition was identical. The NMTRs were corrected to identify the right serial numbers. Although there appeared to be some weaknesses in B&W's quality control processes
in readying fuel for shipment, no concerns were identified with the initial or followup actions taken by the licensee.
l On February 22, while aligning the spent fuel pool (SFP) bridge over the new fuel elevator, the bridge inadvertently " bumped" the west fuel t
transfer mechanism winch platform.
The cause was subsequently e
determined to be an equipment failure of bridge limit switch LS21.
Limit switch LS21 was used to electrically limit bridge movement to prevent certain overtravel conditions. The limit. switch was
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subsequently repaired and the mast and winch inspected for damage.
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addition, it appeared that the operator (non-licensed) manipulating the bridge was.not manually stopping the bridge before reaching travel
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limits but, rather, was relying on limit switches, such as LS21, to stop
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bridge motion in some cases. A senior reactor operator (SRO) licensed individual was in attendance and was in charge of directing the ongoing
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activities at the time. The operations superintendent indicated that l
the operator's actions were not in conformance with operations
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departmental philosophy for manual operation of plant equipment. A Potential Condition Adverse to _ Quality report (PCA193-0058) was
initiated to ensure proper documentation and followup. Followup actions
included suspension of bridge operations during which time the individuals involved in the event were counselled. Subsequently, the
event, as well as management expectations on bridge operations, were
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discussed with all shift crews.
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While reviewing the licensee's procedural requirements associated with
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all three refueling bridges, the inspector could not find any required t
preoperational checks for the SFP bridge. The inspector was informed that there was a preventive maintenance (PM) activity that did the
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prechecks on all three bridges but at the time of the event the
prechecks had not been completed. That apparently was because the two
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bridges in containment were not accessible with the reactor at power and
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the PM was to be performed for all three~ bridges 'at the same time. The licensee also indicated that an informal checkout had been conducted
prior to the SFP bridge being placed into service. The licensee agreed that an assessment into whether more formalized preoperational bridge
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checks should be incorporated was warranted.
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The following day, while moving a new fuel bundle from the new fuel l
storage area to the new fuel elevator using the overhead crane, the
bottom of the fuel assembly " caught' on the upper frame guide of the new
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fuel elevator as it was being lowered into the elevator. This was due
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to improper alignment of the assembly with the elevator. The assembly
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tipped somewhat, and during the recovery operation to lift and right the bundle, it swung approximately 6 inches to the side and impacted on the west fuel transfer mechanism's motor housing. Although the impact knocked some paint chips off the motor housing, no damage was noted to either the motor housing or the fuel assembly.
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The overhead crane was being operated by individuals from the maintenance department under the cognizance of an SRO licensed nuclear
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engineer at the time. Again, as before, work activities associated with fuel movement were suspended and the individuals involved were r
counselled as to their misalignment error.
In this case, as well as in the previous one discussed, licensee management reiterated the need for increased attention to detail while conducting critical work activities.
No violations or deviations were identified in this area.
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Surveillance (61726)
The inspectors observed safety-related surveillance testing and verified that the testing was performed in accordance with adequate procedures,
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that test instrumentation was calibrated, that limiting conditions for 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 other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by
t appropriate management personnel.
The following test activities were observed and/or reviewed:
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DB-MI-03014 Channel Functional Test RPS Channel #4, Reactor Trip Module Trip Logic and Reactor Trip Breaker C.
DB-MI-03204 Channel Functional Test and Calibration of SFRCS Actuation Channel 2, Steam Generator Differential Pressure DB-SC-03071 Emergency Diesel Generator Monthly Test DB-SC-03111 SFAS Channel 2 Functional Test i
DB-SS-03253 Emergency Ventilation System Train #2 18 Month or
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Special Test
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a.
Regarding DB-MI-03014, while performing the required prerequisite steps on January 25, 1993, instrumentation and control (I&C)
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workers observed an abnormal light pattern in the rod control-cabinets indicative of a component. failure. Subsequent
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troubleshooting detected the failed component, and the component
was replaced.
A similar failure in the rod control cabinet resulted in a reactor i
trip on December 12, 1990. The licensee's corrective actions i
resulted in changing testing and operating procedures
prerequisites to ensure that any abnormal light patterns in the j
cabinets would be detected, if present, and that the appropriate
components be repaired prior to system testing.
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The detection and correction of the rod control cabinet anomaly
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I observed on January 25, 1993, was considered a direct result of l
the licensee's corrective action program in that the licensee I'
detected the condition and repaired the affected components prior to it potentially affecting plant operations.
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Regarding DB-SS-03253, on February 5, 1993, the inspector observed i
two engineers tightening charcoal filter modules in the #2
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emergency ventilation system (EVS) filter bank prior to the-conduct of the filter bypass flow portion of the test. The-
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engineers had found that about 20 percent of-the charcoal filter module faceplate bolts and holddown bolts were loose. After-the bolts were torqued in-place, the EVS filter bank access hatch was.
closed and tested using surveillance procedure DB-SS-03253 which
verifies adequate system flow rate and verifies that bypass flow around the filter assemblies is less than 1 percent.
The test was completed satisfactorily.
The inspector questioned the engineers as to how the filter modules' holddown and faceplate bolts became loose and whether.-
these bolts were periodically checked to ensure the filter modules
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were installed in a manner to minimize filtered air bypass flow.
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The engineers believed that the rubber gasketed joints between the j
modules and their housing " relax" in time and that they were not periodically checked. The engineers indicated that the bypass
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flow around the charcoal filters, before tightening, was minimal
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and would not have exceeded the maximum allowable bypass flow of
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1 percent of total flow. This is supported by previous bypass
flow tests which were conducted without bolt tightening and:which
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showed bypass flow was about 0.1 percent. The bypass flow after
the current bolt tightening was about 0.01 percent.
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Technical specification 4.6.5.1.b., required that. EVS testing be performed per Regulatory Position C.5.a of Regulatory Guide 1.52, i
revision 2, which required that a visual inspection of all i
associated components be made before testing. The visual inspection was to be performed in accordance with provisions of L
Section 5 of ANSI Standard N510-1975. The ANSI Standard states l
(in part), "A visual inspection of the components or subsystem to
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be tested, their holding devices, gaskets, housing, and all
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associated components shall be made before each test (experience
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has shown that visual inspection will often reveal deficiencies
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that would otherwise result in failure to pass the test or
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invalidation of its results)... A list of items that should be checked as a minimum, is given in Appendix A."
Appendix A, item A-2, allows for adjustment of nuts to ensure gasket compression of.
50 percent to 80 percent.
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Although the ANSI standard allowed for tightening of joints prior
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to testing, the inspectors were concerned that the preconditioning of equipment prior to testing did not test the equipment in its
"as found" state and as such, a deficient condition would not have
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been identified during the test. This item is cons!dered an
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inspection followup item (346/93004-01(DRP)) pending further NRC i
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review to determine the acceptability of preconditioning the
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aforementioned equipment.
No violations or deviations were identified in this area. One-inspection followup item was identified.
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6.
Maintenance (62703)
Station maintenance activities of safety-related systems and components were observed and/or reviewed during the inspection period to ascertain that they were conducted in accordance with approved procedures, regulatory guides, and industry codes or standards, and in conformance l
with technical specifications.
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The following items were considered during this review:
the limiting 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 performed prior to returning components or systems to service; quality.
t control records were maintained; activities were accomplished by i
qualified personnel; parts and materials used were properly certified;
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radiological controls were implemented; and fire prevention controls were implemented.
Maintenance work orders (MW0s) were reviewed to determine status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which may affect system performance. '
i The following maintenance activities were observed and/or reviewed:
e MWO 2-92-0047-01 Permanent Removal of MS106 Local Control Switch e
MWO 3-93-0923-01 Clean, Inspect, Repair Service Water Pump
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MWO 7-92-0275-04 Replace Service Water Supply Piping to ECCS Room Coolers
MWO 7-92-0463-01 Troubleshoot Safety Features Actuation
System (SFAS) Channel #2
No violations or deviations were identified in this area.
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7.
Manaaement Site Visit and Meetina (30702)
On February 26, 1993, the Director of the Division of Reactor Projects,
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RIII, toured the facility and met with members of the licensee staff..
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Topics discussed were plant performance history, status of plant operations, the upcoming eighth refueling outage, and recent regulatory performance. Afterwards, he addressed an audience of plant workers on recent problems encountered at other plants during outage conditions
including shutdown risk issues.
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Inspection Followup Items
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Inspection followup items are matters which have been discussed with the.
licensee, which will be reviewed further by the inspectors, and which involve some action on the part of NRC or licensee or both. An inspection followup item disclosed during the inspection was discussed in paragraph 5.b.
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Exit Interview The inspectors met with licensee representatives (denoted in paragraph 1) throughout the inspection period and at the conclusion of the inspection on March 15, 1993, and summarized the scope and findings of the inspection activities. The licensee acknowledged the findings.
After discussions with the licensee, the inspectors have determined there was no proprietary data contained in this inspection report.
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