ML20036A326
| ML20036A326 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/05/1993 |
| From: | Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20036A321 | List: |
| References | |
| 50-346-93-06, 50-346-93-6, NUDOCS 9305110090 | |
| Download: ML20036A326 (10) | |
See also: IR 05000346/1993006
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U. S. NUCLEAR REGULATORY COMMISSION
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REGION III
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Report No. 50-346/93006(DRP)
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Docket No. 50-346
Operating License No. NPF-3
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Licensee: Toledo Edison Company
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Edison Plaza, 300 Madison Avenue
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Toledo, OH 43652
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Facility Name: Davis-Besse Nuclear Power Station
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Inspection At: Oak Harbor, Ohio
Inspection Conducted: March 16, 1993, through April 16, 1993
Inspectors:
S. Stasek.
R. K. Walton
J. A. Hopkins
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E. R. Duncan
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Approved By:
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R. D. Lanksbury,(Chief'
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Reactor Projects h ibn 3B
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Inspection Summary
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Inspection on March 16. 1993. throuah April 16. 1993
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(Report No. 50-346/93006(DRP))
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Areas Inspected: A routine safety inspection by resident inspectors of
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operational safety, refueling, surveillances, maintenance, and Temporary
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Instruction 2515/113.
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Results: An executive summary follows:
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Plant Operations:
Overall, performance of the operating crews was good this
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inspection period. . Excellent response to an April 8 trip of decay heat
removal-(DHR) pump #2 was noted (paragraph 2.e).
The pump was being used for
shutdown cooling when it inadvertently ~ tripped. Control room operators
immediately recognized.the pump had tripped and were able.to start DHR pump #1
within~about 8 minutes. Root cause determination will-be followed as an
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inspection followup item.
In general,- adherence to administrative controls
was good. Although the unit was in cold shutdown for refueling throughout the
inspection period with many maintenance and testing activities ongoing,' plant-
housekeeping was maintained at acceptable levels.
Radioloaical Controls: Adherence.to radiation protection program requirements
was good this period with no substantive problems noted.
9305110090 930505
ADOCK 05000346
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Maintenance / Surveillance: Overall, surveillance and maintenance activities
reviewed during this inspection period were conducted in accordance with
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. licensee procedures and regulatory requirements.
Integrated Safety Features
Actuation System testing was well-conducted. However, following completion of
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pressure testing of main steam supply piping to the auxiliary feedwater
system, a wire holding a check valve in the open position was not removed.
This matter is considered an unresolved item pending completion of inspector
review (paragraph.4.b). An abnormal pressure transient occurred on the
service water piping to containment air cooler #1 during surveillance
activities (paragraph 2.a).
The licensee determined that a' temporary piping
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configuration being used during this outage contributed to the event.
Questions which arose following the inspection period will be followed as an
inspection followup item.
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Enoineerino/ Technical Suonort:
Support for outage related activities was good
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during this period. .In particular, engineering provided resources to address-
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.a containment-air cooler (CAC) pressure transient event. Walkdowns were
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conducted of the associated piping and supports, and engineering analyses were
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done to determine specific values of pressures and flows in the subject lines.
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Safety Assessment /0uality Verification:
During inspector review of personnel
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overtime records, some apparent weaknesses were noted relating to the
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administrative control of overtime. However, an assessment of this matter was
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not completed at the conclusion of the inspection period and is considered an
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unresolved item (paragraph 2.g).
Temporary Instruction (TI) 2515/113,
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" Reliable Decay Heat Removal During Outages," was closed.
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Two unresolved items and two inspection followup items were identified during
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this inspection period.
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DETAILS
1.
Persons Contacted
Toledo Edison Company
D. C. Shelton, Vice President, Nuclear
- G. A. Gibbs, Director, Quality Assurance
- L. F. Storz, Plant Manager
- J. W. Rogers, Manager, Maintenance
- S. C. Jain, Director, Engineering
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- E. M. Salowitz, Director, Planning
J. K. Wood, Operations Manager
J. R. Polyak, Manager, Radiological Protection
- V. J. Sodd, Manager, Independent Safety. Engineering
D. R. Timms, Manager, Systems Engineering
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G. M. Grime, Manager, Industrial Security
R. W. Schrauder, Manager, Nuclear Licensing
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C. A. Hawley, Superintendent, Shift Operations
- G. Honma, Supervisor, Licensing
- N. K. Peterson, Engineer, Licensing
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R. C. Zyduck, Manager, Nuclear Engineering
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T. J. Myers, Director, Technical Services
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D. L. Haiman, Manager, Engineering Assurance and Services
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- D. L. Eshelman, Superintendent, Shift Operations
N. L. Bonner, Manager, Design Engineering
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- C. A. Hengge, Supervisor, Systems Engineering
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- T. S. Swim, Supervisor, Design Engineering
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- J. E. Moyers, Manager, Quality Verification
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- L. A. Bonker, Supervisor, Radiation Protection
- T. W. Haberland, Manager, . 0utage/ Planning
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- K. C. Prasad, Staff Engineer, Nuclear Engineering
- Denotes those personnel attending the April 16, 1993, exit meeting.
2.
Operational Safety Verification
(71707) (40500)
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The inspectors observed control room operations, reviewed applicable
logs, and conducted discussions with control room operators during the
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inspection period.
The inspectors verified the operability of selected
emergency systems, reviewed tagout records, and. verified tracking of
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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
hazards,-fluid leaks, and excessive vibrations, and'to verify that
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maintenance requests had been initiated for certain pieces-of equipment
in need of maintenance. Walkdowns of the accessible portions of the
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following. systems were conducted to verify operability by comparing
system lineups with plant drawings, as-built configuration, or present
valve lineup lists; observing equipment conditions that could degrade
performance; and verifying that instrumentation was properly valved,
functioning, and calibrated.
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Station Batteries 2P and 2N
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Emergency Diesel Generators 1-1 and 1-2
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Service Water (safety-related portions only)
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
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:
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a.
On March-19, 1993, during performance of a surveillance, the
portion of the service water system associated with containment
air cooler (CAC) 1-1 experienced an abnormal pressure . transient.
This transient, coupled with pre-existing erosion / corrosion
induced wall thinning, caused a small leak at a pipe elbow on the
CAC. The event occurred following stroking of the_ CAC inlet valve
during the test. The licensee determined that the pressure
transient was caused by the abnormal service water system lineup
during the surveillance which allowed piping downstream of the CAC
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to drain more rapidly than normal, in conjunction with the piping
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configuration on the inlet side to the CAC which allowed some
drainage to occur with the CAC inlet valve closed.
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The licensee walked down affected portions of the service water
system and found that associated pipe supports had not been
damaged and that no additional leaks were present. Although not-
within the scope of the erosion / corrosion program,.the licensee
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was inspecting the CAC system prior to the event and will continue
to monitor performance of the system in the future. At the
conclusion of the inspection period, the licensee had initiated
appropriate procedure changes to prevent future pressure
transients of this type to the service water supply to the CACs.
The leaking CAC pipe bend was replaced and tested satisfactorily.
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Due to the piping arrangements within the CAC, a single piping
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failure of this type at power would not render the CAC inoperable.
In addition, three CACs were installed, while only two are
required. The third CAC functioned as an installed spare.
Based
on these conservative design redundancies, the inspectors did not
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consider this event as safety significant.
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Prior to issuing this inspection report, questions arose
concerning the possibility of a water hammer event following a
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loss of offsite power and the adequacy of corrective actions for
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previous CAC piping failures.
Because these questions could not
be immediately resolved, this matter is considered an inspection
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followup item (346/93006-02(DRP)) pending further inspector
review.
b.
During the inspection period, tours of the auxiliary building
radiologically restricted area (RRA) and containment were
conducted.
Overall, housekeeping was found to be maintained at an
acceptable level and adherence to radiation protection
requirements was good. However, although not overly excessive, an
accumulation of trash and debris was observed. The inspectors
noted that containment cleaning was progressing.
Instances were
also observed where compressed gas cylinder storage practices were
weak.
Specifically, cylinders were observed tied by a very loose
rope to act as the securing mechanism, and multiple cylinders were
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sometimes tied with a common rope. Once brought to the licensee's
attention, these problems were promptly corrected.
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c.
During the outage, the licensee performed eddy current inspections
of all-steam generator tubes in both once through steam generators
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(OTSG) to determine the condition of the tubes.
Based on the
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inspection results, the cost of additional tube inspections and
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other factors, the licensee conservatively decided to plug 321
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tubes in the "A" 0TSG and 43 tubes in the "B" 0TSG. The total
number of tubes plugged was considered small and did not
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significantly alter heat exchange properties of the OTSGs.
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Additionally, the licensee elected to install sleeves in another
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213 tubes in the "B" 0TSG as a preventative measure.
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d.
During ultrasonic inspection activities of the
"B" low pressure
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turbine, the licensee found " indications" on the wheel portion of
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both of the fourth stages. The licensee, upon recommendation by
the vendor, removed the indications by machining the dovetail
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portion of the wheels, removed six blades from the wheel at the
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generator end of the turbine, and removed nine blades from the
wheel at the other end of the turbine. The licensee evaluated and
determined that the turbine could be operated safely in this
condition during the ninth operating cycle with negligible
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reduction in turbine efficiency.
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e.
On April 8, 1993, at 9:22 a.m. (EDT), with reactor vessel water
level at the vessel flange, decay heat removal (DHR) pump #2
tripped due to its electrical power supply breaker opening.
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Control room operators quickly detected that the pump had tripped
and verified that the #1 DHR pump was operable. At 9:30 a.m.,
operators started the #1 DHR pump. Reactor coolant system
temperature increased from 98 F to 105 F during the 8 minutes that
forced circulation cooling of the core was interrupted.
Shortly
after starting the DHR pump #1, reactor coolant system
temperatures returned to their previous values.
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Subsequently, the licensee performed troubleshooting of the
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breaker control circuits and meggered the #2 DHR pump motor and
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found no deficiencies. The breaker was replaced with a spare
breaker, and the system was declared operable and placed in
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standby at 12:33 a.m. on April 9,1993.
A similar event occurred with the same breaker on
September 17, 1991, (reference inspection report 346/91017(DRP)).
During the earlier event, the #2 DHR pump supply breaker had also
inadvertently opened. An investigation by the licensee indicated
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the most likely cause of the trip was an inadvertent bumping of
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the control switch, since work was being performed in an adjacent
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cabinet. However, a confirmation could not be made through
personnel interviews conducted at the time.
At the end of this inspection period, the licensee was continuing
to investigate the cause of the second event.
Because the subject.
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breaker had tripped on two occasions and the root cause had not
been determined, this matter is considered an inspection followup
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item (346/93006-01(DRP)) pending further inspector review of the
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licensee's final root cause determination and associated followup
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actions.
f.
On April 13, 1993, at 2:10 p.m., with the plant in Mode 5, while
lining up decay heat train #1 to recirculate the borated water
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storage tank (BWST), control room operators noted that reactor
coolant system inventory was decreasing as indicated by a decrease
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in pressurizer level. The operators then closed DH1517, the DHR
pump #1 suction valve, which terminated the event.
Decay heat train #2 continued to operate throughout the event and
was not affected by the reduction in inventory. This event will
be the subject of a special inspection (reference inspection
report 346/93011(DRP)).
g.
The inspectors reviewed licensee staff and contractor personnel
overtime records for the month of March 1993, to determine if
plant management was controlling the use of overtime as required
by Technical Specification 6.2.3.
For the records reviewed,
several apparent inconsistencies were found where individuals
appeared to have exceeded administrative limits without the
required management authorization.
However, further review was
needed at the conclusion of the inspection to determine if this
was indeed the case. Therefore, this matter is considered an
unresolved item (346/93006-02(DRP) pending completion of the
inspectors' followup.
h.
On April 15, 1993, at about 3:30 p.m. (EDT), with the plant in
Mode 5, the #2 Emergency Diesel: Generator (EDG) was declared
inoperable due to the moisture content in the fuel oil day tank
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which exceeded the chemistry limit of 0.05 percent.
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The licensee was making preparations to perform Integrated Safety
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Features Actuation System (SFAS) time response testing when'
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chemistry samples taken from the #2 EDG fuel oil day tank
indicated that the moisture content was 0.22 percent. The #2 EDG
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was then declared inoperable.
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Both the #2 EDG day tank and the #2 EDG week tank.had been drained
March 23, 1993, for a cleanliness inspection, then refilled.
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- 2 EDG was started twice after the maintenance activity was
completed and operated without problems. The licensee purged the
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day tank sample line and resampled the day tank. These were found
to be within the required limits. The #2 EDG was returned to
operable status on April 15, 1993, at 8:45 p.m.
The licensee was
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continuing to investigate how the moisture entered the tank at the
end of the inspection period. The inspectors will follow the
licensee's evaluation and further corrective actions (if needed)
as part of the routine inspection program.
Technical Specifications required that in Mode 5 one EDG with its
support systems be operable.
throughout this period.
No violations or deviations were identified in this area.
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3.
Refuelina
(607101
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During this inspection period, the plant remained shutdown for the
eighth refueling outage. The inspectors observed portions of fuel
reload and vessel reassembly activities.
Procedures for reload,
including prerequisites, were reviewed and verification of plant
conditions were made to verify compliance with Technical Specifications.
All inprocess activities observed appeared to have been conducted in
accordance with the licensee's procedures and within regulatory
requirements.
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a.
On April 3,1993, following placement of the reactor vessel head
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on the reactor vessel but prior to tensioning, an inadvertent
water addition was made to the reactor vessel. This resulted in
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vessel water level rising above the vessel flange, and seepage
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past the vessel head o-ring seals and into the vessel stud holes.
This necessitated cleaning of the bolt holes.
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This event will be reviewed as part of a special inspection
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(reference inspection report 346/93011(DRP)).
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b.
Following completion of core reload activities, the inspectors
viewed the licensee's videotape of fuel assembly serial number
verification and insured that the fuel had been placed in
. conformance with the licensee's approved cycle 9 core map.
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,
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
Specifications 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
appropriate management personnel.
The following test activities were observed and/or reviewed:
DB-ME-03002
Station Battery Service and Performance Discharge Test
DB-MI-03007
Time Response Test of Reactor Protective System
Channel #3
DB-PF-03065
Pressure Tests
DB-PC-10117
Station Blackout Diesel Generator Blackout Load Test
DB 9 *'114
Safety Features Actuation System Integrated Time
Response Test
DB-a
3710
Control Room Emergency Ventilation System Train #1
18-Month Surveillance Test
DB-SC-04026
13.8 KV Bus "A"
& "A" Fast Transfer Test
DB-SC-04053
C1/C2 Bus Lockout Test
a.
The inspectors witnessed the performance of portions of
surveillance test DB-SC-03114, Safety Features Actuation System
(SFAS) Integrated Time Response Test and noted that the test was
well-executed.
Personnel involved with the test were well-briefed
prior to starting the test. The brief included contingency plans
if problems arose, the proper use of communications, and control
room chain of command. The licensee also provided additional
management resources during this infrequently performed test to
provide oversight and to ensure a high level of safety was
maintained. The inspectors noted that procedures used by
personnel were current and equipment used during the test was
calibrated. An initial review of the test data revealed that all
major equipment required to operate during the test functioned as
designed.
b.
During the performance of DB-PF-03065, pressure test of the main
steam supply piping to the auxiliary feedwater pump turbines, test
personnel held check valve MS735 in the open position with a wire.
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After the test was completed, however, the wire holding the check
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valve open was not removed. The installation of the wire was not
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documented as part of the test and contributed to its not being
removed after the test was completed.
The inspectors consider
this to be an unresolved item (346/93006-03(DRP)) pending further
review.
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No violations or deviations were identified in this area.
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Maintenance (62703)
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Station maintenance activities of safety-related systems and components
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were observed and/or reviewed during the inspection period to ensure
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that they were conducted in accordance with approved procedures,
regulatory guides, and industry codes or standards, and in conformance
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with Technical Specifications.
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The following items were considered during this review:
the limiting
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conditions for operation (LCO) were met while components or systems were
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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
control records were maintained; activities were accomplished by
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qualified personnel; parts and materials used were properly certified;
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radiological controls were implemented; and fire prevention controls
were implemented.
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Maintenance work orders (M40s) were reviewed to determine status of
outstanding jobs and to assure that priority was assigned to safety-
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related equipment maintenance which may affect system performance.
The following maintenance activities were observed and/or reviewed:
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Maintenance on #1 EDG
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Maintenance on #2 EDG
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Troubleshoot and Test Inverter YV1 (MWO 7-90-0640-04)
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Installation of New Turbine Bypass Valves (MWO 2-92-0012-XX)
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Moisture Separator Reheat Piping Replacement (MWO 2-90-3026-XX)
Troubleshoot Source Range Nuclear Instrument #2 (MWO l-93-0344-00)
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Troubleshoot Reactor Protection System Channel #3
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(MWO 7-93-0176-01)
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Auxiliary Feedwater Steam Admission Valve Installation
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(MWO 2-87-1273-04)
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Refurbishment of #2 Service Water Pump (MWO 3-93-09?.3-01)
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Reactor Coolant Pump 1-2-1 Uncoupled Run (MWO l-91-1535-21)
No violations or deviations were identified in this area.
6.
Review of Temocrary Instruction (TI) 2515/113
(C1csed) TI 2515/113. Reliable Decay Heat Removal. During Outages. The
inspectors reviewed the licensee's shutdown risk program to ensure
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reliable decay heat removal capability could be maintained during plant
outages.
The licensee issued NG-PS-00116, " Outage Nuclear Safety Control," which
established specific guidelines relating to the scheduling' of outage
activities and the availability of plant systems, structures,.and
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components necessary to ensure electrical power, decay heat removal, and
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other plant functions.were available to ensure that an adequate safety
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margin existed during plant shutdown. This procedure incorporated.the
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guidance of NUMARC 91-06 and INPO 92-02 documents dealing with issues
related to reduced reactor vessel inventory during shutdown. The
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licensee organized a multi-disciplinary outage review group to review
the eighth refueling outage schedule and determine how it would impact
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shutdown safety issues. The group concluded that the outage schedule
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maintained an acceptable margin of safety and defense-in-depth to guard
against shutdown risk concerns.
The inspectors determined that the licensee's shutdown risk program was
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fully implemented and consistent with guidance documents. The
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licensee's approach to shutdown risk issues was conservative with
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contingency plans available.
Independent Safety Engineering Group
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(ISEG) personnel monitored plant conditions and provided periodic status
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reports addressing important shutdown risk parameters. Temporary
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Instruction 2515/113 is closed.
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No violations or deviations were identified in this area.
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7.
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.
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During this inspection,. unresolved items were identified in paragraphs
2.g and 4.b.
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8.
Inspection Followuo Items
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Inspection followup items are matters which have been discussed with the
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licensee, which will be reviewed further by the inspectors, and which
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involve some action on the part of the NRC, or the licensee, or both.
Inspection followup items-disclosed during the inspection are discussed
in paragraphs 2.a and 2.e.
9.
Exit Interview
The inspectors met with licensee representatives (denoted in
paragraph 1) throughout the inspection period and at the conclusion of
the inspection on April 16,1993, and summarized the scope and findings
of the inspection activities. The licensee acknowledged the findings.
After discussions with the licensee, the inspectors determined there was
no proprietary information contained in this inspection report.
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