ML20135C260
| ML20135C260 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 11/15/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20135C258 | List: |
| References | |
| 50-413-96-16, 50-414-96-16, NUDOCS 9612060251 | |
| Download: ML20135C260 (21) | |
See also: IR 05000413/1996016
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION II
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Docket Nos:
50-413, 50-414
License Nos:
Report Nos.:
50-413/96-16. 50-414/96-16
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Licensee:
Duke Pcwer Company
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Facility:
Catawba Nuclear Station. Units 1 and 2
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Location:
422 South Church Street
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Charlotte. NC 28242
Dates:
September 8 - October 19, 1996
Inspectors:
R. J. Freudenberger, Senior Resident Inspector
P. A. Balmain, Resident Inspector
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R. E. Carroll, Project Engineer. Region II
R. L. Franovich Resident Inspector
C. W. Rapp, Senior Reactor Inspector, Region II
Approved by:
L'. D. Wert, Acting Chief
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Reactor Projects Branch 1
Division of Reactor Projects
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ENCLOSURE
9612060251 961115
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ADOCK 05000413
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EXECUTIVE SUMMARY
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Catawba Nuclear Station. Units 1 & 2
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NRC Inspection Report 50-413/96-16, 50-414/96-16
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This integrated inspection included aspects of licensee operations
maintenance, engineering, and plant support. The report covers.a 6-week
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period of resident ins)ection; in addition, it includes the results of
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announced inspections ay regional reactor safety and reactor projects
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inspectors,
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Ooerations
Unit 1 restart observations were generally positive in the areas of
containment cleanliness. the conduct of zero power physics testing, and
the licensee's resolution of an intermediate range reactor trip setpoint
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discrepancy (Section 01.1).
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Because of ineffective communications between the operations and
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chemistry organizations. Unit 1 entered Technical Specification (TS)
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3.5.4.b action statement when fueling water storage tank boron
concentration slightly exceeded its TS maximum limit (Section 01.2).
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Minimal entries in operations logs precluded their use as a diagnostic
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tool that may have led to the earlier isolation of a reactor coolant
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system filter leak on Unit 1 (Section M1.1).
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Maintenance
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Non-Cited Violation 50-413/96-16-01 was identified because an inadequate
leak test resulted in leakage from a reactor coolant system letdown
purification filter not being identified and subsequent contamination of
the 560 foot level of the Auxiliary Building and some areas of the 543'
foot level (Section M1.1).
Unit 1 restart was appropriately delayed to evaluate the cause of
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pressurizer pressure control problems and its impact on safe operation
of the facility (Section M1.2).
The licensee identified that the surveillance test for determining
controlled reactor coolant system leakage rate was not being performed
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in accordance with the TS basis for the test.
The actions to correct
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the specific deficiency were timely and appropriate.
Pending review of
long-term corrective actions, this issue was identified as Unresolved
Item 50-413.414/96-16-02 (Section M3.1).
Testing of control rods performed during the Unit 1 outage complied with
NRC Bulletin 96-01 Control Rod Insertion Problems (Section M8.1).
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ENCLOSURE
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Executive Summary
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Enaineerina
The licensee adequately addressed two issues associated with the Service
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Water System Operational Performance Inspection (Sections E8.2 and
E8.4).
Two other issues from the same inspection remain open: (1)
Violation 50-413.414/94-17-02 remained open pending NRC review of
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thermal performance modeling, and (2)
Inspector Followup Item 50-
413.414/94-17-10 remained open pending additional radiographs to be
performed on the nuclear service water to auxiliary feedwater line since
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records of a previous radiograph could not be located (Sections E8.1 and
E8.3).
Plant Sucoort
An isolated case of an inattentive fire watch was identified by NRC
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inspectors.
Licensee followup actions were appropriate (Section F4.1).
The delay in correction of a probable ficor drain system clog prior to a
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reactor coolant system filter leak reflected a lack of operational focus
within the chemistry organization (Section M1.1).
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ENCLOSURE
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Reoort Details
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Summary of Plant Status
Unit 1 was in a refueling / steam generator replacement outage until October 2.
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when unit restart commenced,
The unit outage ended on October 4, when the
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. reactor entered Mode 1.
The unit reached full power on October 10 and
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operated at or near 100% power for the remainder of the inspection period.
Unit 2 operated at or near 100% power for the duration of the inspection
period.
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Review of UFSAR Commitments
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A recent discovery of a licensee operating their facility in a manner contrary
to the Updated Final Safety Analysis Report (UFSAR) description signified the
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need for a special focus review that compares plant practices, procedures.
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and/or parameters to the UFSAR descriptions. While performing inspections
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discussed in this report, the inspectors reviewed the applicable portions of
the UFSAR that related to the areas ins)ected.
The inspectors verified that
the UFSAR wording was consistent with cle observed plant practices.
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procedures, and/or parameters.
No deficiencies were identified.
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I. Ooerations
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Conduct of Operations
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01.1 Unit 1 Restart Observations (61726. 71707. 40500)
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Containment Cleanliness Walkdown
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The inspector conducted two cleanliness tours of the reactor building
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and pipe chase perimeter before the unit entered Mode 4.
The inspector
noted during the first tour, performed on Seatember 21. that a number of
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items had not been removed from the reactor Juilding or secured to
structures.
Water covered 'most of the containment floor at that time.
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as flushing was in progress in the overhead to wash loose articles and
debris to the containment floor for easy retrieval and removal.
The inspector entered containment again on September 22. The
containment floor was dry and much cleaner, but several items (i.e.,
nails. wire, a hand towel, and a pipe end) were identified and carried
out of the building.
The inspector also made note of some material
condition discrepancies and communicated those items to the Reactor
Building Coordinator.
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The above housekeeping tours were conducted prior to the licensee's
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final Mode 4 closecut. The inspector found the condition of the reactor
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building to be much improved from the first tour and considered the
controls for containment cleanliness to be effective.
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ENCLOSURE
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Loose Parts Monitor Indications Durina HeatuD
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On September 26, the licensee suspended Unit 1 heatup at approximately
420 degrees F because of frequent loose parts monitor alarms in the
reactor vessel head area.
The licensee utilized the data acquisition
and retrieval features of the recently installed loose parts monitoring
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system to determine that the alarms resulted from thermal expansion of
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piping in the vicinity of the loop B hot leg. An industry expert
reviewed the data and determined that no reactor coolant system damage
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resulted because of the small magnitude of the thermal expansion events.
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Based on these evaluations the licensee resumed Unit 1 heatup. The
inspector considered these actions an example of effective engineering
support to operations.
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Zero Power Physics Testino
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The inspector attended the prejob briefing for zero power physics
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testing.
The briefing handout provided a detailed plan for the testing,
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discussion of applicable requirements, and clearly delineated
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responsibilities. The briefing was conducted in the control room.
However, the high noise level created by the control room ventilation
system in conjunction with testing and maintenance activities in the
vicinity of the briefing, caused multiple distractions.
For these
reasons, the designated management lead for the infrequently performed
testing requested that the briefing be repeated in a conference room
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outside the control room, where the noise level was lower and the
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environment was more conducive to successful communication.
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inspector considered this action to be indicative of conservative and
conscientious management oversight.
The inspector observed portions of zero power physics testing. The
licensee's im)lementation of several improvements to the test program
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continued to 3e effective. These included locating test equipment in
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the control room horseshoe area and face to face communications between
test personnel, control room supervisors and operators. The inspector
observed an example of conservative actions when testing was suspended
to repair a cable problem with an information only reactivity chart
recorder located at the " operator at the controls" desk. The inspector
reviewed data collected and results for the isothermal moderator
temperature coefficient measurement (PT/0/A/4150/12A).
The increased
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heat transfer characteristics of the new Unit 1 steam generators were
evident to the operators performing the test. The inspector verified
that the reactor coolant temperature and core reactivity changes were of
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sufficient magnitude to generate acceptable test results. The inspector
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concluded that testing was well coordinated and controlled.
Evaluation of Intermediate Rance Trio Setooint Discreoancy
The inspector reviewed the licensee's resolution of a reactor trip
setpoint discrepancy associated with intermediate range channel N35.
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ENCLOSURE
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Technical Specification (TS) Table 2.2-1 indicates an intermediate range
trip setpoint of 5 25% Rated. Thermal Power (RTP), with an allowable
value of 5 31% RTP. The purpose of this trip is to provide core
protection during reactor startup to mitigate the consequences of an
uncontrolled rod withdrawal from a subcritical condition-
Power range
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low setpoint trip also provides redundant protection to the intermediate
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range trips.
During a setpoint verification performed at approximately 20% power per
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PT/0/A/4150/01. Controlling Procedure for Startup Physics Testing,
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licensee estimated that the N35 trip setpoint would have exceeded the TS
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allowable limit of 31%. With N35 blocked, measurements of current
values were collected as )ower approached 31%.
The licensee
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subsequently determined tlat the actual trip setpoint of N35 was greater
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than 25% RTP. but was less than the TS allowable limit of 31% RTP.
From discussions with engineering personnel, the inspector discerned
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that a slight asymmetry in core power contributed to the N35 setpoint
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3roblem. The intermediate range setpoints are developed and calibrated
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3ased on the assumption of uniform core aower. Actual core power in the
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area adjacent to detector N35 was less tlan anticipated, which resulted
in an apparent low setpoint when compared to total core calorimetric
power.
The inspector determined that the licensee took appropriate
actions to declare N35 inoperable (PIP 1-C96-2700), and verified that
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intermediate range N36 and power range (low) channels were within the TS
trip setpoint of 5 25% RTP.
The licensee continues to investigate the
cause of the core power tilt and anticipates that it will be reduced
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later in core life.
The inspector concluded that the licensee took
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appropriate actions to identify and evaluate the N35 trip setpoint
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discrepancy.
01.2 Fuelina Water Storaae Tank Boron Concentration
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Insoection Scooe (71707)
On October 2, with Unit 1 in Mode 2 during zero power physics testing,
the weekly sample of boron concentration in the fueling water storage
tank (FWST) revealed that boron concentration was in excess of the upper
limit referenced in TS 3.5.4.b and specified in the Core Operating
Limits Report (COLR).
Zero power physics testing was temporarily
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suspended. demineralized water was transferred to the FWST. and boron
concentration was returned to its allowable limits within the time
period of TS 3.5.4.
The inspector discussed the issue with licensee
personnel and reviewed the TS the TS Basis, the COLR. and test results
from previous samples during the past two months.
b.
Observations and Findinas
The COLR specifies a boron conrentration lower limit of 2475 ppm and an
u)per limit of 2575 ppm in modes 1 through 4.
The inspector determined
tlat boron concentration had been as high as 2644 on September 14 after
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ENCLOSURE
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highly borated water from the refueling cavity had been transferred to
the FWST following the completion of refueling operations.
Demineralized water makeup to the FWST reduced boron concentration to
2494 ppm on September 16.
Between September 16 and 19. FWST level
dropped by roughly 30% for reactor coolant system fill and vent.
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Blended makeup to the FWST was calculated to result in a final boron
concentration of 2525 ppm. the limit mid-range.
However, on September
20, boron concentration was 2570 ppm and fluctuated between that
concentration and 2558 ppm until September 25.
On October 2. boron
concentration reached 2579 ppm. and Unit 1 entered TS action 3.5.4 to
restore boron concentration within allowable limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be
in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
The licensee took immediate corrective actions to lower FWST level by
transferring inventory to the spent fuel pool so that a calculated
dilution could be initiated.
Once the dilution was completed, the FWST
was recirculated via the A containment spray pump and another sample was
obtained.
Boron concentration had dro
ed to 2559, and FWST operability
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wasrestoredwithinthe6-hoursallowad)ebyTS.
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Considering the length of time that boron concentration was close to the
u)per limit and the o)portunity to reduce boron concentration to avoid
tie potential inopera)ility of the FWST. the inspector questioned the
effectiveness of communications between the Operations and chemistry
organizations. The inspector discussed the issue with plant personnel
and determined that, although the chemistry staff had notified the
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control room that FWST boron concentration was high and needed to be
reduced, operations personnel did not recognize the operational impact
of any further baron concentration increase.
The licensee determined that operations personnel did not exercise a
questioning attitude to ensure that they understood how boron
concentration affected FWST operability, and that the chemistry
organization did effectively communicate the consequences of failing to
reduce FWST boron concentration. To effect im)rovement in the interface
between these two organizations, the licensee las designed several
initiatives to: (1) hold daily meetings between the two groups.
(2) designate an o)erations point of contact for chemistry concerns.
(3) draft status sleets for monitoring critical chemistry parameters and
communicating items with immediate or impending impact to plant status
during the daily operations meeting.
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Conclusions
The inspector concluded that this issue was indicative of ineffective
communications between chemistry and operations.
A second example of
interface issues between chemistry and operations is discussed in
section M1.1 of this inspection report.
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ENCLOSURE
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Operations Organization and Administration
06.1' Administrative Control of Keys
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a.
Insoection Scooe (71707)
On September 4. the inspector signed out keys for access to the Nuclear
Service Water System pumphouse and the switchyard.
The key issued from
the Work Control Center to access the pumphouse was not the proper key
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to unlock the pumphouse door.
The keys for the switchyard were correct.
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The inspector informed the licensee.
The licensee initiated PIP C96-
2399 to address the discrepancy. The inspector reviewed the PIP
corrective actions.
b.
Observations and Findinos
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Until recently, either of two keys would open the pum3 house door. The
lock sets on the Technical Support Center (TSC) and t1e 0)erations
Support Center (OSC) had been changed to make one of the ceys unique to
them. When this change had been made, the key log was updated to
reflect the correct keying for the TSC and OSC. but the entry for the
pumphouse was not updated to reflect that only one key would now fit the
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pumphouse door.
The cause of the problem was identified as' incorrect updating of the key
log. The licensee performed an audit of the key log and found no other
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discrepancies.
In addition, a change to the key log updating
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was initiated which included identification and evaluation of ceys that
have duplicate entries in the log.
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Conclusions
The inspector considered the error in the key log to be an isolated
instance. The licensee took appropriate corrective action.
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II. Maintenance
M1
Conduct of Maintenance
M1.1 Letdown Purification System Leak in Auxiliary Buildina
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a.
Insoection Scooe (62703. 40500)
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On September 7.1996. Unit 1 core reload was delayed due to leakage in
the letdown purification loop which was in service for refueling cavity
purification. The leakage, which originated from the letdown 1B Reactor
Coolant (NC) filter that had been put in service, contaminated the 560
foot level of the Auxiliary Building and some areas of the 543 foot
level. The inspector reviewed the circumstances surrounding the event.
ENCLOSURE
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assessed recovery actions. and evaluated the licensee's Failure
Investigation Process (FIP).
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Observations and Findinas
The estimated 20 gpm leak originated from the IB NC filter, which is
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between the letdown ion exchangers and Volume Control Tank (VCT) in the
Unit 1 Chemical Volume Control System.
Since the NC filter is put in
service remotely because of its location in a covered pit. the filter
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housing-to-cover leak that was draining to the 560 foot floor drain
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system went undetected when placed in service around 2:30 a.m. on
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September 7.
Approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later a portion of the 560 foot
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floor drains were found to be backing up, and they continued to do so
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until letdown was isolated at approximately 8:52 a.m.
Letdown NC Filter 1B Leak
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Based on a review of completed Enclosure 4.4 of OP/1/A/6200/01.
Transferring From NC filter 1B to 1A. the inspector confirmed that the
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18 NC filter had been placed in service for a leak test on Se]tember 3.
1996. following a filter cartridge change out.
After the leac test, the
filter was isolated /placed in standby, where it remained until being
placed in service on September 7.
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From photographs taken of the "as found" condition of the IB NC filter
cartridge, the inspector noted the existence of a slight downward bend
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Jart way around its top o-ring mounting lip (perhaps caused when the
linged filter housing cover was closed).
The photographs also revealed
that the grooved o-ring pulled away from the mounting lip directly
opposite the bend.
In view of these "as found" conditions, the
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inspector verified through record review that charging and letdown from
the Residual Heat Removal (RHR) System were in progress during the leak
test on September 3.
Subsequently, in order to fully understand how the
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IB NC filter passed its leak test the inspector requested a plot of VCT
level for the time in question.
Although not conclusive, the plot
showed a relatively acute downward slope corresponding to the short time
(approximately 11 minutes) that the 1B NC filter was in service on
September 3.
A closer review of the aforementioned Enclosure 4.4
indicated that once the filter was valved into service, the leak test
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was performed in a very short period of time (approximately 1 minute).
In view of the short duration of this leak test. the fact that it was
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performed (like the cartridge replacement) from above looking down into
the pit, and that the pit cover (as indicated in Enclosure 4.4 ) was
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installed approximately one minute later, it is conceivable that the
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filter housing-to-cover leakage indicated by the VCT level plot went
undetected.
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To preclude such problems in the future, the licensee incorporated a
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filter leak test in associated Maintenance Procedure MP/0/A/7150/060.
Pall-Trinity Filter Removal and Replacement, which requires the filter
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to be under pressure for at least 10 minutes before inspecting for
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leaks.
This change reflects the ASME functional test requirements for
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class II (B) and III (C) systems att recuired to operate during normal
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31 ant operation. The inspector reviewec the revised procedure and.
laving confirmed that it is used on all such filters in the plant, found
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it appropriate. Additionally, the licensee implemented actions to-
determine if other maintenance procedures require similar changes. This
licensee-identified and corrected violation is characterized as Non-
Cited Violation 50-413/96-16-01. Inadequate Reactor Coolant Filter Leak
Test, consistent with Section VII.B.1 of the NRC Enforcement Policy.
560 Foot Floor Drain System
The cause of the 560 foot floor drain system backup was subsequently
found to be blockage caused by a ball of small diameter rope, several
welding rod stubs, some tie wraps, and approximately 5 gallons of resin.
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The inspector reviewed Problem Investigation Process (PIP) Report 0-C96-
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1795 (dated July 15. 1996) concerning the discovery of a large amount of
resin in a pre-strainer of the subject floor drain system. as well as
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higher than usual dose levels on associated inlet piping.
Screened as
non-significant. PIP resolution remained with chemistry to clean out the
strainer.
PIP updates approximately one month later indicate that inlet
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pipe dose levels hadn't dropped significantly, but actions to remove the
resin internal to the drain system would be postponed until September
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after U1E0C9 was completed. Since the higher than usual inlet pipe dose
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levels were an indication of blockage (i.e. , a barrier where resin was
accumulating behind). the inspector considered the postponement of
internal resin removal to be indicative of a lack of operational focus
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on the impact of potential drainage increases to the floor drain system.
Recovery of Affected Eauioment
Some of the water which backed up the 560 foot floor drain system made
its way to a number of the pump rooms on the 543 foot level by seeping
through non-water tight hatch covers / floor plugs.
Equipment exposed to
water and wetted included a Unit 2 spent fuel cooling system panel,
three environmentally qualified Rotork valves (2NI-135B.136B. and
100B). and motors associated with the 1A and 1B charging pumps and the
2B and 1B safety injection pumps.
The inspector reviewed data from pump
motor / cable testing performed per IP/0/A/4974/13. Horizontal Split
Sleeve Bearing Motor Inspection and Maintenance, and PT/0/A/4950/01.
Power Cable Testing, as well as the work requests documenting the
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results of water intrusion inspections, to confirm that no equipment
problems resulted from the spill.
A review of FSAR Section 3.3.6.1.
Critical Auxiliary Building Areas, indicated that equipment submergence,
not wetting, is the concern from flooding in the Auxillary Building.
Since submergence of critical Auxiliary Building areas was not a concern
in this event, there was no significant risk of flood damage to safety
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equipment.
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A general tour by the inspector of the 560. 543. and 522 foot levels of
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the Auxiliary Building revealed no additional cencerns.
No personnel
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contaminations occurred as a result of the floor drain backup or the
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decontamination efforts to restore access to the affected areas.
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Leak Isolation /00erator Response
Unit 1 had entered Mode 6 at approximately 3:45 a.m., on September 7.
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According to the event time line developed by the licensee's FIP, the _
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Unit 1 night shift control room crew suspected reactor coolant system
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leakage early on. but VCT level trending was difficult (due to charging
flow indication problems and the necessity to balance charging and
letdown to maintain VCT level high to compensate for an inability to
provide gas overpressure).
The control room operators took the
following actions: monitored containment floor and equipment-sumps:
dispatched an operator to verify sperit fuel cooling system purification,
since it had been put in service earlier in the shift: dispatched an
operator (based on misleading information from Radwaste Chemistry
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personnel) to ensure VCT divert valve NV-172A wasn't leaking back to the
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recycle holdup tank since being placed in automatic earlier in the
shift: referred to AP/1/A/5500/26. Loss of Refueling Canal or Spent Fuel
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Pool Level, but didn't enter because specific symptoms were not met; and
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pursued the spent fuel cooling system demineralizer as a possible leak
source since it was placed in service during the shift and resin was
found in the floor drain strainer.
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The inspector reviewed the Unit 1 Supervisor and Control Room Operator
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logs for the time of interest.
Although information was limited
concerning the above activities, the Unit 1 supervisor log did contain
two entries around 5:00 a.m. concerning notification of the 560 foot
floor drain backup and suspected reactor coolant system leakage. The
Control Room Operator log revealed no indication of a problem, nor
mentioned the control room actions discussed above. Although placing
the IB NC filter in service around 2:30 a.m. had been prompted by the
control room due to a high differential pressure on the 1A NC filter,
there was no indication of such in either log until an end of the shift
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control room operator log entry at 6:19 a.m. documenting issuance of R&R
16-2099 to replace the 1A NC filter.
Noting that the night shift placed
the 1B NC filter in service and finally having a quantified leak rate
(based on VCT level) of approximately 20 gpm the day shift
subsecuently: suspended core alterations: entered AP/1/A/5500/26: and
closec 1KF-122 (cavity to spent fuel pool cross connect) and secured
purification and charging / letdown.
Based on a review of the actions taken, the inspector determined that
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the operators acted approariately to the information readily at hand.
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It was also apparent to t1e inspector that the lack of sufficient
information regarding system / equipment problems and operational status
changes (e.g. , makeup volumes. NC filter realignment, etc. . .) precluded
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the use of the Unit 1 Supervisor and Control Room Operator logs as an
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effective diagnostic tool.
Such a tool may have led to isolating the
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leak earlier and reducing the area contaminated.
[a4 lure Investiaation Process (FIP) Review
A review of the licensee's FIP for this ' event found'it to be
appropriately thorough.
It addressed and provided proposed corrective
actions for such related issues as: the validity of using existing
prestrainers in the 560 foot floor drain system to collect particulate
materials that the downstream oil and grit removal tank is designed to
collect: Radwaste Chemistry needing to ensure all conse
considered when establishing the time frame for a plan'quences ares implementatio
I
communication deficiencies between Operations and Chemistry which
prevented the transfer of complete and accurate information: Operations
referral to (versus entry into) AP/1/5500/26 and whether additional
guidance is needed for smaller magnitude leaks: the encountered
workarounds which made trending VCT level difficult; and whether
existing criteria for log entries are reflecting Management's
expectations. As discussed earlier. the FIP also tasked Maintenance
with ensuring applicable test requirements are clearly stated in the
necessary procedures.
c.
Conclusions
The licensee's FIP and recovery actions for the NC filter leak and
subsequent Auxiliary Building floor drain backup were considered to be
appropriate.
Post-cartridge replacement leak testing was inadequate to
l
ensure the leak tightness of NC filter 18. Subsequent implementation of
a minimal hold period at test 3ressure should preclude recurrence. The
i
postponement of removing a pro]able floor drain clog prior to the leak
reflected a lack of operational focus within the Chemistry organization.
Minimal entries in operations logs precluded their use as a diagnostic
tool that may have led to an earlier 1 solation of the leak.
M1.2 Pressurizer Pressure Control Problems Durina Unit 1 Restart
a.
Insoection Scoce (62703)
During Unit I restart from the steam generator replacement outage.
control room operators were unable to maintain pressurizer pressure with
the normal C bank of pressurizer heaters. A leaking pressurizer spray
valve was suspected. and inspection and repair activities revealed that
both pressurizer spray valves (INC-27 and 1NC-29 from reactor coolant
system cold legs A and B. respectively) were leaking.
The inspector
discussed the issue with licensee personnel
reviewed the station
Problem Investigation Process (PIP) Report that documented the problem.
reviewed work orders, and evaluated the licensee's decision to continue
unit restart with one of the valves still leaking.
l
ENCLOSURE
.
.
10
b.
Observations and Findinas
On September 27. Unit 1 was in mode 3 at normal operating pressure and
temperature, preparing for mode 2 operation.
Control room operators
noticed that pressurizer pressure control could not be maintained
,
without the additional heat input from the backup pressurizer heater
banks. The suspected cause was leakage past a spray control valve.
'
Work order 96077934 was initiated to inspect and repair 1NC-27. The
licensee determined that the valve positioner was out of adjustment and
proceeded to calibrate the positioner and declare the valve operable.
When the unit was preparing to enter mode 2 on Seatember 30 the problem
with maintaining pressurizer pressure without baccup heaters was
encountered again. Troubleshooting revealed that leakage past INC-29
was the cause.
Work order 96078640 was initiated on October 1 to
ins)ect and repair INC-29. The valve positioner and actuator appeared
to 3e correctly adjusted, and seat leakage was determined to be a result
of some internal degradation.
On October 2. the licensee decided to
continue with the unit restart and initiate work orders to perform more
extensive work on both pressurizer spray valves in the next refueling
outage.
The inspector reviewed PIP 1-C96-2673.
The licensee determined that
numerous leakage and setup problems had been experienced on all
pressurizer spray valves. At the end of the inspection period, the need
j
for predefined work orders on each valve was being evaluated by the
Engineering organization.
No other concerns were identified.
c.
Conclusions
I
The inspector concluded that the licensee appropriately delayed Unit 1
startup to evaluate the cause of the pressurizer pressure control
problem and its impact on safe operation of the facility. The decision
to continue the Unit 1 restart with leakage past 1NC-29 was adequately
justified.
M3
Maintenance Procedures and Documentation
M3.1 Nonconservative Reactor Coolant System Controlled Leakaae Test
a.
Insoection Scooe (617_251
On September 26. the licensee discovered that the surveillance test for
determining controlled Reactor Coolant (NC) System leakage rate was not
conservative. The ins)ector discussed the finding with plant personnel
and reviewed the TS, t7e corrected procedure, and the associated station
Problem Investigation Process (PIP) report.
ENCLOSURE
l
11
l
b.
Observations and Findinos
During a review of the proposed Im3 roved Technical Specification (ITS)
'
3.5.5.1, the licensee discovered t1at the PT/1(2)/A/4150/01, NC System
'
Controlled Leakage Verification, was not being performed to simulate the
system flowpath as it is described in the current TS basis.
Specifically, the basis states that the controlled leakage limitation
restricts operation when the total flow supplied to the reactor coolant
pump seals exceeds 40 gpm with the modulating valve in the supply line
(NV-294) fully open at a nominal NC System pressure of 2235 psig. This
limitation ensures that in the event of a loss of coolant accident
'
(LOCA), the safety injection flow will not be less than assumed in the
safety analyses.
The surveillance test had been performed with NV-294 in the normal
modulating position to control charging flow. This was not conservative
because the accident analysis assumes a station blackout concurrent with
the LOCA. and the valve fails to the open position on a loss of power to
ensure that adequate seal injection is provided.
)
Once the discrepancy was identified the licensee determined that the
surveillance for Unit 2 had been missed as a result of the discrepancy.
Licensee Event Report (LER) 413/96-09 is currently being drafted to
document the procedural inadequacy, and the licensee's past operability
evaluation will be included in the report.
The licensee initiated an
,
immediate procedure change to ensure that the surveillance could be
performed correctly within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed by TS 4.0.3.
Unit
1 was in Mode 3 with NC System pressure at 1900 psig; the surveillance
was not required for that unit until it reached normal operating
pressure (2235 psig).
The surveillance was performed in both units on September 27.
The
controlled leakage for Unit 1 was 35'.5 gpm; the controlled leakage for
Unit 2 was 33 gpm.
,
c.
Conclusions
The inspector concluded that the licensee was proactive in identifying
the discreaancy in the test procedure and correcting it in a timely
manner.
T1e impact of the procedural error was minimal, and the
subsequent test results indicated that controlled leakage with the
modulating valve fully open remained less than the 40 gpm limit imposed
by TS.
Pending a review of the licensee's past operability evaluation
and long-term corrective actions, this issue will be identified as
Unresolved Item (URI) 50-413.414/96-16-02:
Nonconservative RCS
Controlled Leakage Test.
ENCLOSURE
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12
M8
Miscellaneous Maintenance Issues (92902)
M8.1 (Closed) Insoector Followuo Item (IFI) 50-413/96-08-03: Review of
End-Of-Cycle (EOC) Control Rod Drop Timing Data.
Control Rod Insertion Problems, reported that control rods had failed to
fully insert in fuel assemblies with greater than 30.000 Megawatt
Days / Metric Tonne Uranium (MWD /MTU) exposure and requested measurement
and evaluation of drag forces for all fuel bundles with control rod
assemblies.
In accordance with Bulletin 96-01, the licensee had conducted control
rod assembly drag tests of fuel bundles with control rod assemblies at
EOC 9.
The inspector reviewed fuel bundle drag testing traces and
identified ten fuel bundles that indicated increased drag during control
rod insertion.
Of these ten fuel bundles. the licensee had identified
four fuel bundles as having longer control rod drop times at EOC
compared to Beginning-of-Cycle.
The inspector reviewed the list of fuel
bundles discharged at EOC 9 and determined that only one of the ten fuel
I
bundles was reloaded for Cycle 10 and that fuel bundle did not have a
control rod assembly.
Based on the ins)ectors review, the licensee had
conducted E0C drag testing as required )y Bulletin 96-01 and identified
those fuel bundles that may be susceptible to excessive friction.
No
further licensee action for Bulletin 96-01 was required.
Nine fuel bundles, discharged at E0C 8, were reinserted for Cycle 10
with five of these fuel bundles having control rod assemblies. The
inspector noted that Bulletin 96-01 did not address the reinsertion of
fuel bundles other than those off-loaded at E0C and questioned if these
reinserted Cycle 8 fuel bundles had been drag tested.
The licensee said
only the fuel bundles from EOC 9 had been drag tested.
The licensee
said they could take data during drag testing conducted after the upper
internals package was set.
However, this drag testing was conducted
!
only to assure the control rod drive and the control rod assembly were
connected and would not yield the same detailed traces obtained during
EOC 9 drag testing. The EOC 10 exposure for the reinserted Cycle 8 fuel
bundles was predicted to be greater than the maximum exposure for fuel
bundles drag tested during E0C 9.
The inspector questioned if the
i
j
licensee had considered this factor since the E0C 9 drag testing would
!
not be bounding. The licensee stated that this was not considered a
aroblem because Bulletin 96-01 addressed problems observed only with
destinghouse Vantage-5 fuel and not Framatome Mark-BW fuel used in both
Catawba Units 1 and 2.
The inspector noted that Bulletin 96-01 did not
distinguish between different fuel manufacturers' fuel bundles.
However, the inspector concluded that the drag testing conducted af' c
setting the upper internals package was adequate to meet the
requirements of Bulletin 96-01.
.
!
ENCLOSURE
!
13
III. Encineerina
E8
Miscellaneous Engineering Issues (92903)
E8.1
(Ocen) Violation (VIO) 50-413.414/94-17-02:
Failure to Properly
Translate Regulatory Requirements into Specifications. Drawings, and
Procedures.
The licensee had revised calculation CNC-1150.01-00-001.
Standby Nuclear Service Water Pond - Thermal Analysis During One Unit
LOCA and One Unit Shutdown, revision 6. to irclude groundwater recharge
and seepage from the Standby Nuclear Service water Pond (SNSWP) dam.
The licensee calculated the total inventory loss for the 30-day duration
to be 7.36 ac-ft due to groundwater recharge and 0.007 ac-ft due to
seepage from the SNSWP dam.
However, the license only used inventory
losses for the first six days as the input to the thermal analysis
model. The licensee had calculated the peak service water intake
temperature occurred on the 5th day after event initiation.
One
additional day of heat input was added for conservatism to obtain the
maximum service water intake tem3erature. Calculation CNC-1150.01-00-001
did not reference any analysis t1at validated using only 6-day inventory
losses verses 30-day inventory losses.
The licensee provided a
sensitivity analysis that used the 30-day inventory losses as an input
to the model.
This sensitivity analysis indicated that the additional
inventory losses did not affect service water intake temperature beyond
the 6-day inventory losses used in the thermal analysis model.
The licensee had also revised CNC-1150.01-00-001 to include pump work as
a heat in)ut to the model.
However, this input was substantially less
than the leat input from various other auxiliary heat inputs, such as
motor and oiler coolers, already included in the thermal analysis model.
Based on a review of the calculation and supporting documentation, the
inspector concluded the licensee had adequately addressed SNSWP
inventory losses due to seepage and increased heat input due to pump
,
work.
However, this violation will remain open pending completion of
I
additional NRC review of the SNSWP modeling.
EB.2 (Closed) VIO 50-413.414/94-17-05:
Failure to Perform Quality Related
Activities per Prescribed Procedures or Drawings.
The inspector toured
the Service Water (RN) pumphouse and found that the licensee had
corrected the instrument lines having a ]roblem as identified in
Inspection Report (IR) 413.414/96-10.
T1e licensee had identified that
one of the instrumentation lines needed a supporting tray installed and
l
had written a deficiency.
The licensee had also placed warning signs
against climbing on the instrument lines.
The inspector concluded the
licensee's actions were adequate.
As identified in IR 413.414/96-10. procedure SI/0/A/5090/001. Tube
Fitting and Tubing Installation. revision 0. was not clear as to which
section in Enclosure 4.1 was referred to in Enclosure 4.6.
The licensee
had considered this a generic problem with the format of the Standard
ENCLOSURE
l
l
14
Procedures being developed and issued PIP 96-2449 to initiate a review
of the format for these procedures.
The inspector concluded this review
would address the concern.
E8.3 (Ocen) IFI 50-413.414/94-17-10:
Flush Program Improvements.
The
inspector had previously requested to review the radiographs of the RN
i
supply to Auxiliary Feedwater (CA) line, but the licensee was unable to
produce these radiographs during the earlier inspection. The licensee
i
attempted to locate these radiographs, but was unable to locate the
'
radiographs for the A train RN supply to CA line.
Furthermore, there
was no documentation of the as-found condition for the A train RN supply
to CA line on the work order.
The licensee had issued a PIP to initiate
documentation of the as-found conditions and to radiograph the RN su) ply
to CA line.
This item will be reviewed after the radiographs are tacen.
The licensee had scheduled this activity for late 1996.
E8.4 (Closed) IFI 50-413.414/94-17-12:
Unnecessary Post-Maintenance Tests
(PMTs) on Predefined Work Orders.
The licensee had previously conducted
a review of the 3 redefined work orders and identified the unnecessary
PMTs listed in tie predefined work orders.
However, the licensee had
not updated the predefined work orders to reflect only the required
PMTs.
The inspector reviewed a sample of the predefined work orders for
the RN system and found they had been updated.
IV. Plant Sucoort
F4
Fire Protection Staff Training and Qualification
F4.1
a.
Insoection Scoce (71750. 40500)
On September 4. during a facility tour, the inspector identified that an
individual performing fire watch duties was not attentive to those
duties. The licensee was informed of the inspector's observation. The
inspector reviewed PIP C96-2955 and the licensee's followup actions.
b.
Observations and Findinas
A floor plug which was part of a fire barrier had been removed for
equipment access during the Unit 1 outage.
As a compensatory action, a
fire watch was posted at the opening.
The inspector observed the
individual performing fire watch duties reclined, with his shoes and
socks removed. When the inspector left the area, the individual was on
his feet and attentive. The inspector informed the licensee of his
observation.
The same day, an individual assessment was performed by
the licensee's safety review group.
The assessment included a visit to
all areas with fire watches )osted (7 total) .
The same individual was
found inattentive again by t1e licensee.
ENCLOSURE
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15
F
l
The licensee took disciplinary action with the individual and
j
implemented refresher training to reinforce fire watch responsibilities
and expectations based on lessons learned.
c.
Conclusions
.
Based on the results of the licensee's assessment and other facility
tours performed by the ins)ector. the inspector considered the
inattentive fire watch to 3e an isolated instance.
The licensee took
appropriate corrective action.
'
V. Manaaement Meetinas
X1
Exit Meeting Summary
The inspectors ) resented the inspection results to members of licensee
management at t1e conclusion of the inspection on October 29. 1996. The
licensee acknowledged the findings presented.
No proprietary information was
identi fied.
l
,
ENCLOSURE
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16
PARTIAL LIST OF PERSONS CONTACTED
Licensee
Bhatnager. A.. Operations Superintendent
Coy. S., Radiation Protection Manager
Forbes, J., Engineering Manager
Harrall . T.
IAE Maintenance Superintendent
Kelly. C., Maintenance Manager
Kimball . D. , Safety Review Group Manager
Kitlan, M., Regulatory Compliance Manager
Lowery, J.. Compliance Specialist
McCollum, W.. Catawba Site Vice-President
Nicholson, K., Compliance Specialist
Patrick, M., Safety Assurance Manager
Peterson. G., Station Manager
Propst, R., Chemistry Manager
Rogers
D. , Mechanical Maintenance Manager
Tower
D., Compliance Engineer
~
l
ENCLOSURE
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_
17
,
i
'
INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
i
IP 40500:
Effectiveness of Problem Identification and Prevention
IP 61726:
Surveillance Observation
IP 62703:
Maintenance Observation
IP 71707:
Plant Operations
IP 71750:
Plant Support Activities
IP 92902:
Followup - Maintenance
IP 92903:
Followup - Engineering
ITEMS OPENED. CLOSED, AND DISCUSSED
Ooened
50-413/96-16-01
Inadequate Reactor Coolant Filter Leak Test
(Section M1.1).59-413.414/96-16-02
Nonconservative RCS Controlled Leakage Test
(Section M3.1)
Closed
50-413/96-08-03
IFI
Review of E0C Control Rod Drop Timing Data
(Section M8.1).
l
50-413.414/94-17-05
Failure to Perform Quality Related Activities
Jer Prescribed Procedures or Drawings (Section
E8.2).
!
50-413,414/94-17-12
IFI
Unnecessary PMTs on Predefined Work Orders
.(Section E8.4).
J
Discussed
50-413.414/94-17-02
Failure to Properly Translate Regulatory
'
Requirements into Specifications. Drawings, and
Procedures (Section E8.1).
50-413.414/94-17-10
IFI
Flush Program Improvements (Section E8.3).
50-413/96-09
LER
Inadequate NC System Controlled Leakage
Verification (Section M3.1).
ENCLOSURE
1
18-
LIST OF ACRONYMS USED
ASME -
American Society of Mechanical Engineers
CA
-
-
Beginning-of-Cycle
,
CFR
-
Code of Federal. Regulations
COLR -
Core Operation Limits Report
DEV
-
Deviation
-
Duke Power Company
E0C
End-of-Cycle
-
-
Failure Investigation Process
FSAR -
Final Safety Analysis Report
FWST
Fueling Water Storage Tank
-
IAE
-
Instrument and Electrical
i
IEEE -
Institute of Electrical and Electronic Engineers
IFI
-
Inspector Followup Item
IR
-
Inspection Report
-
Improved Technical Specification
KF'
-
Spent Fuel Pool Cooling System
LER
-
Licensee Event Reporr.
,
LOCA -
Loss of Coolant Accident
-
Motor Control Centers
MWD /
-
Megawatt Days / Metric Tonne Uranium
NC
-
i
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Non Cited Violation
i
1
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Operations Support Center
-
' Problem Investigation Process
PMTs -
Post-Maintenance Tests
Parts per million
-
RHT
-
Recycle Holdup Tank
-RN
Nuclear Service Water System
-
Rated Thermal Power
-
SNSWP -
Standby Nuclear Service Water Pond
TS~
-
Technical Specifications
'
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UFSAR -
Updated Final Safety Analysis Report
-
Unresolved Item
Vac
-
Volts alternating-current
-
Volume Control Tank
'
Vdc
-
Volts direct-current
-
Violation
-
Work Order
ENCLOSURE
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