ML20045D175
| ML20045D175 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 06/01/1993 |
| From: | Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20045D171 | List: |
| References | |
| 50-458-93-10, NUDOCS 9306280057 | |
| Download: ML20045D175 (18) | |
See also: IR 05000458/1993010
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APPENDIX
U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-458/93-10
Operating Li anse: NPF-47
Licensee:
Gulf States Utilities
P.O. Box 220
St. Francisville, Louisiana 70775-0220
Facility Name:
River Bend Station
Inspection At:
St. Francisville, Louisiana
Inspection Conducted: March 14 through April 24, 1993
Inspectors:
W. F. Smith, Senior Resident Inspector
D. P. Loveless, Resident Inspector
R. H. Bernhard, Senior Resident Inspector,
Grand Gulf Nuclear Station
Approved:
k
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J.E.~GagBardo7ChTef,PyectSectionC
Date
Inspection Summary
Areas Inspected:
Routine, unannounced inspection of plant status, onsite
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response to events, operational safety verification, maintenance and
surveillance observations, and an engineered safety feature walkdown.
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Results:
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The licensee's actions in the early plant staff recognition and
identification and the reactor engineering analyses, planning, and
implementation of testing to locate and suppress further degradation of
the reactor fuel leak were' excellent, with one exception, as described
below (Section 2.1).
The operators departed from the licensee's self-checking process when
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they scrammed the wrong control rod during reactor flux tilting tests.
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A noncited violation was identified in' view of the minimal safety
significance of the event and corrective actions taken (Section 2.1).
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The licensee's corrective actions in conducting detailed reviews and
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identifying and correcting overlap discrepancies in logic system
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functional test surveillance procedures continued to demonstrate good
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performance. - However, the number of discrepancies confirmed the need to
perform the detailed reviews (Section 2.2).
Finding a test fixture installed in Source Range Monitor (SRM) A after
completion of the functional test and documentation of its removal was-
identified as a potential violation pending further review to determine
the cause. An unresolved item was opened (Section 2.3).
The licensee's response to insure that an occurrence similar to the
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Perry Nuclear Plant emergency core cooling system (ECCS) suction
strainer clogging event should not occur at River Bend Station was
thorough and timely (Section 2.4).
Control room operator performance and professionalism was found to be
improving, based on routine daily observations and observation of the
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shutdown that was implemented on April 17 and 18, 1993 (Section 3.1).
The performance of maintenance observed during this inspection period
was good.
Procedures provided by the maintenance work orders (MW0s)
appeared to be appropriate in each case.
Summar_y of Inspection Findings:
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A noncited violation was identified (Section 2.1).
Unresolved item 458/93010-1 was opened (Section 2.3).
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Inspection Followup Item 458/93010-2 was opened (Section 5.1)-,
Attachment:
Attachment - Persons Contacted and Exit Meeting
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DETAILS
1 PLANT STATUS
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At the beginning of this inspection period, the plant was operating at
100 percent power.
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On March 26, 1993, the licensee reduced reactor power to 65 percent, in order
to perform flux tilt testing.
This testing is further described in
Section 2.1 of this report.
The reactor was returned to 100 percent power on
April 1.
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On April 2, a problem with the turbine moisture separator reheaters resulted
in a power reduction to 92 percent, until April 5 when power was restored to
100 percent after repairs.
Power remained at 100 percent until April 17, when the plant was shut down and
cooled to ambient temperature and pressure to replace reactor recirculation
pump seals.
The outage was originally scheduled to end on April 28.
However,
because a main steam isolation valve failed to close, the outage was extended
to identify the root cause(s) and to implement corrective actions.
Details of
the main steam isolation valve problems were documented in special NRC
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Inspection Report 50-458/93-18.
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At the end of this inspection period, the plant remained shut down at; ambient
conditions.
2 ONSITE RESPONSE TO EVENTS (93702,61726)
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2.1 Reactor Fuel Integrity Failure
As documented in NRC Inspection Report 50-458/9:-05, on February 25, 1993, the
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licensee identified a fuel cladding failure as ividenced by a significant
increase in the offgas pretreatment release rate.
On March 23, the licensee
requested relief from the required actions of Technical Specification 3.1.4.2,
" Rod Pattern Control System," to allow for continuous rod withdrawal of an
individual rod to its original notch position after the rod was fully inserted
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by a manual scram from a partially inserted position.
The purpose of the
relief request was to allow performance of testing to help. locate the leaking
fuel rod.
On March 26, the NRC granted the relief in a Notice of Enforcement Discretion,
which concluded that plant safety would not be unduly compromised and that
such action was in the best interest of public health and safety.
The relief
was granted for a period of 7 consecutive days, beginning from the first use
of the relief from the Technical- Specification requirements.
Later.that day, the licensee reduced reactor power to 65 percent and began
testing. The inspectors observed portions of the testing as it was conducted
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in accordance with Test Procedure TP-93-0006, Revision 0, " Flux Tilting." The
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inspectors reviewed the procedure.to ensure it reflected the relief as
described in the Notice of Enforcement Discretion and that it was otherwise
consistent with the facility license. 'The procedure was.well structured for.
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an orderly progression of the' test and had appropriate' prerequisites,
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precautions, limitations, and independent verifications.
However, General
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Guidelines Step 6.4, and Procedure Step 9.1.7 provided-direction to
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continuously insert, rat M r than scram, control rods-that were not fully
withdrawn.
The inspectors noted that the Notice of Enfurcement Discretion was
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issued with the understanding that the relief from Technical Specification
requirements was necessary to allow scramming of control rods that were not
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fully withdrawn and to allow continuous withdrawal of the control rod to
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return it to its pretest condition.
Upon discussing the natter with cognizant _
NRC staff, the inspectors confirmed this understanding.
The licensee was requested to supplement their request for relief prior to
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proceeding in a manner that was contrary to the methods described in the
Notice of Enforcement: Discretion. 'On March 29, the licensee' submitted a
letter providing appropriate technical and safety ' justification. On March 30,
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the NRC staff gave the licensee verbal concurrence to proceed with testing
partially inserted control rods by inserting rather than scramming the rods.
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Even though the licensee failed to clearly communicate the methodology
intended for testing partially inserted con +rol rods to the NRC, the
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inspectors and the NRC staff considered the amecded method to be more
conservative, in that 'it reduced the severity of mechanical and hydraulic
stresses on the control rods and control rod drives.
The NRC staff followed
up with written amendment to the Notice of Enforcement Discretion on_
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April 8, 1993.
As specified in Procedure TP-93-0006, testing of the 120 fully withdrawn
control rods was accomplished by Surveillance Test Procedure STP-052-3701,
Revision 7, " Control Rod Scram Testing." This procedure had been used in the
past to conduct scram timing tests of 10 percent of the control rods once per
120 days of operation in o'rder to satisfy Technical Specification surveillance
requirements for operability. .This provided a well-established, contro11ed'
method of inserting and withdrawing individual rods for the : flux tilting test.
On March 29 an apparent breakdown in communications between two operators at
the hydraulic control units resulted in scramming the wrong control rod.
The
procedure required Rod 20-25 to be scrammed, but the operator scrammed
Rod 20-45 instead. The licensee identified and documented the error on
Condition Report 93-0152.
Intended Rod 20-25 was in the fully withdrawn
position.
Rod 20-45 was in the. fully inserted position, so there was no rod-
movement and no unexpected flux changes. The hydraulic control unit for
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Rod 20-45 was restored to its proper lineup and testing was halted. After-
counseling the operators on proper communications.and self-checking, the shift
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supervisor directed that a copy of the control rod movement' sheet'from
Procedure TP-93-0006 be provided to the operators at the hydraulic control
units so that they could visually compare the rod to be tested, with the
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verbal test- direction coming from the control room. The remainder of. testing
was completed without any additional errors.
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The-licensee conducted an evaluation to determine the facts,. event chronology,
-and any existing conditions pertaining to the event in accordance with
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Procedure OSP-018, Revision 1, " Operations Accountability Review." The
results of that review revealed a change in the self-checking routine used by
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the same operators on all other rods. . In this instance, one operator relied
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on the other to select the next hydraulic control unit based on his
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overhearing a telecommunications repeat-back, and then the~first operator did
not visually verify, by the label plate, that he was scramming the correct
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rod.
The licensee also took personnel actions appropriate to the
circumstances.
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Failure to scram the correct control rod in the sequence established by
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Procedure TP-93-0006 was in violation of Technical Specification 6.8.1.d,
which requires test activitie.s of safety-related equipment to be implemented
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by written procedures.
However, since the' licensee's corrective actions were
appropriate, this licensee-identified violation is not being cited because the
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criteria specified in paragraph VII.B.2 of Appendix C to 10 CFR Part.2 of the
- NRC's " Rules of Practice" were satisfied.
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By March 31, the licensee had identified the approximate location of the fuel
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leak. When Rod 44-29 was inserted from 40 notches to 00 notches', there~was a.
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significantly greater change in radiation levels in-the offgas pretreatment-
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piping as read on a temporarily installed Eberline SRM 200 radiation monitor.
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In addition, as the operators attempted to withdraw the rod back to its
original position, radiation levels increased at such a rate that the reactor
engineer directed the rod to be reinserted, and left inserted, to. suppress
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neutron flux in that part of the core.
After retesting three rods in close proximity to Rod 44-29, the ;icensee
confirmed that dod 44-29 was closest to the fuel leak and concluded that
operation for the remainder of the fuel cycle would be with Rod 44-29 inserted
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to minimize the local power in the vicinity of the leak.
In that manner,
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degradation of the leak would probably be minimized while at power.
The flux tilting process used by the licensee appeared to be successful in
identifying the location of the fuel leak so that power and consequent
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degradation could be suppressed by leaving the appropriate rod' inserted.
Offgas, grab samples taken on April 6, after full- power operation was resumed
and stabilized, indicated approximately 8,000 microcuries per second.
Prior
to testing, offgas grab samples indicated approximately 12,000 microcuries per..
second.
Because the recoil levels before the leak was noted were about
4,000 microcuries per second, the release rate was cut approximately -in half.
Despite'the operator error in scramming the wrong rod and the poor
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communications between the NRC staff and the licensee over the testing ~ methods
to be used, the licensee's performance in the identification, reactor
engineering analyses, planning, and implementation of the testing to find and
suppress further degradation of the reactor fuel leak was nottworthy. The
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inspectors noted good support of the test by the radiation protection
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organization by providing a sensitive and responsive radiation monitor and
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recorder setup to detect changes in offgas radiation levels.
This
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significantly reduced exposures that would have been received through
otherwise frequent grab sampling.
2.2 Continued Discrepancies Found in Logic System Functional Tests (LSFTs)
In Section 2.1 of NRC Inspection Report 50-458/93-05, the inspector described
licensee-identified discrepancies found in certain LSFT surveillance test
procedures.
In each case, the problem was a failure to maintain testing
continuity or overlap on LSFTs to assure that all logic components, from the
sensor through the actuated device, were tested to verify operability.
The
licensee's initiative in finding the deficiencies and taking appropriate
immediate, as well as long-term corrective actions, was appropriate.
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addition, the components in question were found to be operable in each case.
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During this inspection period, as a result of long-term corrective action
accelerated because of the above findings, the licensee identified the
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following additional deficiencies in LSFTs:
On March 26, the licensee identified that a wire connecting the reactor
core isolation cooling (RCIC) low supply pressure contacts to the
remainder of the logic circuit was not being tested.
Also the wire
connecting the high main steam line temperature contacts to the
remainder of the RCIC logic was not being tested.
Vacuum breaker
isolation Valves E51-F077 and E51-F078 required an RCIC low supply
pressure signal, among others, to automatically isolate.
One of the
wires connecting the appropriate relay contacts was not tested by any
procedure.
Technical Specification 4.0.3 was implemented, and the
surveillance test procedures were changed and subsequently performed.
The circuits were all verified to be operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This was
documented in Condition Report (CR) 93-0145.
On March 26, the licensee identified a condition in the safety relief
valve / automatic depressurization system LSFT where no check was made to
verify that the "LO-LO SET" setpoint remained in effect until the reset
pushbutton was pressed as required by Technical Specification 4.4.2.2.1.b.
Technical Specification 4.0.3 was
implemented, the test procedures were revised, and the circuits were
verified operable within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
This was documented in CR 93-0145.
On March 31, the licensee identified that there was no test in the
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Division I and Il electrical power systems LSFT to verify the function
of the contacts associated with the loss of coolant accident signal,
which was in series with the short time degraded voltage signal.
The
contacts were tested and verified operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This was
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documented in CR 93-0158.
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On April 5, the licensee found the contact used to trip Standby _ Service
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Water Pump ISL?*P2C on receipt of a sustained undervoltage signal' on the
Division 111 bus was not being tested.
The procedure was revised and
the contact was tested and found to be operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. .This
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was documented in CR 93-0168.
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On April 12, the licensee identified that the Division III electrical
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power system LSFT did not test to verify the function of the contacts
associated with the loss of coolant accident signal that was in series
with the short-time degraded voltage signal,
This condition was similar
to those of Divisions I and 11 discussed above found on March 31.
The
procedure was corrected and the contact was tested and found to be
operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This was dor umented in CR 93-0178.
On April 15, the licensee discovered that there was no monthly channel
functional test on two relays associated with the RCIC high; steam flow
timer.
Surveillance test procedures were revised and implemented within
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the circuits were found to be operable. This was
documented in CR 93-0184.
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At the end of this inspection period, the licensee was approximately
35 percent complete on their detailed review of LSFTs.
On March 15, 1993, the
licensee documented these issues in Licensee Event Report (LER)93-002, with a
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commitment to update the report. On April 2, the license issued a supplement-
to the LER with a continuing commitment to update the LER incrementally as
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additional LSFT overlap problems were discovered as a result of their
corrective action.
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In all cases, the untested circuits were found to be operable. Additional
inspection will be conducted as followup to LER 93-002 to review the
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licensee's completed reviews.
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2.3 Test Fixture Found in SRM A
On April 17, 1993, while performing postshutdown source range nuclear
instrument functional surveillance testing on SRM A, a test fixture used to
hold down the inoperative inhibit pushbutton-was found installed.
The
technicians were performing Surveillance Test Procedure STP 503-4503,
Revision 6A, " Control Rod Block-Source Range Monitor A Weekly Channel
Functional (C51-K600A, C51-R601A, C51-R602A)." When the " operate" switch was
taken out of the " operate" position in accordance with Section 7.1.7 of the
procedure, the appropriate annunciators failed to alarm as expected. Upon
opening the instrument drawer, the technicians found the test fixture
installed. The licensee documented the. deficiency on CR 93-0196.
The inspector reviewed the completed surveillance test documentation from the
last time SRM A was tested, which was on December 7, 1992, during the previous
outage.
Section 7.2.2 of Procedure STP-503-4503 required removal of the
fixture, thus releasing the pushbutton.
This section was signed off as
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compl eted .
There was also a signed-off independent verification in
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Attachment 4, with Step 2 verifying restoration of the pushbutton_and removal
of the test fixture.
The licensee could not produce any subsequent
documentation showing that SRM A was tested or worked on since December.7,
1992, nor was it apparent to the inspector that anyone would have a need to
install the test fixture until April 17. As of the end of this_ inspection
period, the licensee did not have a cause identified to explain why the test
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fixture was installed.
The inspector questioned the proper implementation of
the licensee's independent verification program.
Previously, the inspector
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identified a possible weakness in this program in NRC Inspaction
Report 50-458/92-32, Section 5.1.
The licensee's investigation was not
completed and, therefore, appropriate corrective actions had not yet been
taken.
The safety significance of the test fixture being left installed was mitigated
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by the fact that, when the instrument was operating, the function switch had
contacts which were closed in parallel with the inoperative inhibit
pushbutton.
Therefore, the instrument was capable of performing its intended
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safety functions when it was required to be operable.
The inspector was
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concerned, however, that this event indicated a possible breath in the
integrity of the licensee's independent verification program or a failure to
control the installation of the test fixture, if it was reinstalled after
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December 7, 1992.
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Either scenario is a potential violation of NRC regulations, pending the
results of further review. This is an unresolved item (458/93010-1).
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2.4 Licensee Response to Perry Strainer Issue
On April 15, 1993, a potential common mode failure was identified at the Perry
Facility, a General Electric boiling water reactor (BWR-6) with a Mark III
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containment.
The suppression pool suction strainer for Residual Heat Removal
Pump B was observed to have accumulated substantial amounts of debris.
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February 1993, they cleaned the suppression pool and replaced the residual
heat removal suction strainers following identification of strainer fouling
and deformation.
NRC Region IV staff informed Gulf States Utilities of the
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event and expressed concern that similar circumstances might apply to
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River Bend Station, given that River Bend Station did not have an installed
suppression pool cleanup system.
The licensee responded by assembling an ECCS suction strainer task force
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comprised of a senior management sponsor, a team leader from system
engineering management, and representatives from system engineering, design
engineering, operations, and chemistry, with licensing and quality assurance
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oversight.
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The inspector monitored the licensee's. activities and reviewed the results of
actions taken. Activities included:
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A review of pump inservice data and trends.
An operational events search for similar events.
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A design review and comparison between River Bend Station and Perry
Station considering strainer construction and location of suction and
discharge points.
Suppression pool cleanliness and sampling plans.
(During the 1992
refueling outage, River Bend emptied and cleaned the suppression. pool.)
Divers to inspect strainers, suppression pool, and sample bottom of pool
for precipitates or sludge.
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Conduct of resin blockage tests on strainers with smaller holes.
In addition to the above tasks, before disturbing the installed ECCS strainers
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and' dispatching divers, the licensee conducted a 24-hour test run of the low
pressure core spray pump at full design flow, with return to the suppression
pool.
The suction strainer was not cleaned prior'to the test.
The test was
run in accordance with special Test Procedure TP-93-0009, Revision 0, "ECCS-
Suction Strainer Test Procedure." The inspector witnessed portions of the
test and reviewed the data. The test procedure implemented the normal pump
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inservice test surveillance procedure with an extended run to obtain data.
After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, there was no degradation of suction pressure, as read on a
precision test gauge.
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The ECCS suction strainers at River Bend Station were found to be different
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from the perforated cone type sheet metal strainers at Perry Station. The
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River Bend Station strainers were made up of a series of perforated plates and
spacers bolted together.
It appeared that they would not collapse, even if
subjected to a full vacuum on the pump suction side.
However, the perforated
plates had 3/32-inch holes which could become clogged.
The strainers were
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manufactured by Zurn Industries and were designed to provide adequate flow
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with 50 percent clogging.
The licensee's ECCS Suction Strainer Task Force had not reached any final
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conclusions as of the end of this inspection period, but they informed the
inspector that the following items were under review, pending results of the
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above listed actions, to prevent an event similar to Perry Station's problem
from occurring at River Bend Station:
Chemistry sampling frequency
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Strainer inspection frequency
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Inservice test procedures
Specific trending for suction strainers
Tests and sampling after safety relief valve actuations
Overall, the inspectors concluded that the licensee's response to this
potentially significant issue was timely, thorough, and proactive.
The
inspectors will continue to monitor the licensee's activities on this issue
through completion of final actions.
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2.5 Conclusions
The early plant staff recognition and identification and the reactor
engineering analyses, planning, and implementation of testing to locate and
suppress further degradation of the reactor fuel leak were excellent, with one
exception as described below.
A noncited violation was identified for operators departing from the
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licensee's self-checking process when they scrammed the wrong control rod
during reactor flux tilting tests.
The licensee's corrective actions in seeking out, identifying,.and correcting
overlap discrepancies in LSFT surveillance procedures continued to demonstrate
good performance. The number of discrepancies confirmed the need to perform
the detailed reviews.
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Finding a test fixture installed in SRM A after completion of' the functional
test was a potential violation pending further review to determine the cause.
An unresolved item was opened.
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The licensee's response to ensure that an occarrence similar to the Perry
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Station ECCS suction strainer clogging event should not occur at River Bend
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Station was thorough and timely.
3 OPERATIONAL SAFETY VERIFICATION (71707)
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The objectives of this inspection were to ensure that this facility was being
operated safely and in conformance with regulatory requirements and to ensure
that the licensee's management controls were effectively discharging the
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licensee's responsibilities for continued safe operation.
3.1 Control Room Observations
The inspectors observed operations in the control room on a daily basis when
on site.
Specific notice was taken of the licensee's work control process,
during the outage, to replace recirculation pump seals and other selected
work. The inspectors noted that implementation of a recent change to' isolate-
the work control station from the immediate area of the shift supervisor was
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working well_ and eliminated a significant amount of distraction from ~the. shift
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-supervisor and_the control operating foreman.
The licensee had placed- a
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senior reactor operator in charge of the work control. station who was? always
in close communication with the control room and was knowledgeable of the
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types of information the control room needed.
The inspectors noted that overall operator professionalism in the control room
during power operation and when shut down was good.
Shift turnover appeared
to be thorough and informative, and communications-observed between the
operators and control room supervisors was generally crisp and concise.
On April 17 and 18, 1993, the inspectors observed portions of the plant shut
down for Planned Outage 93-01.
The purpose of the outage was to replace
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degrading reactor recirculation pump mechanical seals.
The plant was reduced'
from full power at about 10 percent per hour.
Before the- planned scram from
about 30 percent power, the shift supervisor conducted a detailed briefing on-
what his expectations were from each operator.
The condensate storage tank.(CST) was essentially isolated because-of high-
total organic concentrations (discussed in NRC Inspection
Report 50-458/93-05). Therefore, the shift supervisor made sure that the
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operators were in a position to anticipate the effects of a different makeup:
water lineup.
Specifically, reference was made to the suction lineup for the
control rod drive hydraulic (CRD) pumps. The suction pressure regulator from
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the condensate header was. set abnormally low to minimize return flow to the
CST. Also, the CRD pump miniflow lines, which returned to the CST, were
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temporarily isolated. An operator was stationed'at the CRD pumps to restore
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suction and miniflow in the event of a pump trip when the control- rod -
hydraulic cylinders refill after the scram.
The scram was executed from 29 percent power at 3:19 p.m.
The operators were
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well coordinated and maintained reactor vessel level within the expected band.
The operating CRD Pump A tripped as anticipated on low suction pressure.
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operators experienced some difficulty in restoring CRD makeup flow, ostensibly
because of the slow response of the pressure regulator.
By 3:49 p.m. CRD pump
flow was restored. Overall, the shutdown to ambient conditions was safely
conducted.
3.2 Plant Tours
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On March 24, 1993, during a routine tour, the inspector observed that the
inner door of the 171-foot elevation containment personnel airlock was not
locked in a manner that would prevent the door from being opened.
As
described in NRC Inspection Report 50-458/93-11, the NRC has determined that
the airlock interlock mechanism was inoperable.
Technical Specification 3.6.1.4,-Action b, requires, with,the interlock mechanism
inoperable, that one airlock door be locked closed at all times.
During
personnel entry and exit through'the airlock, an individual must be dedicated
to assure that two doors are not opened simultaneously.
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Although the licensee considered the airlock to be operable at'the time, the
NRC staff disagreed.
This was established during a conference call between
Region IV personnel, NRR personnel, the resident inspectors, and licensee
personnel on March 11,-1993. The Plant Manager issued a directive to
implement the requirements of Technical Specification 3.6.1.4, Action
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Statement b, as a conservative action.
The inspector had noted.that this'
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action placed the plant in compliance with the Technical Specifications.
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However, during the March 24 tour, the inspector observed the firewatch' unlock
and open the outer door.
Following the outer door being chained, the
inspector observed a radiation protection technician.open the inner door and
enter containment without the benefit of the key. 'The technician stated'that
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he had obtained prior permission from the firewatch and that.the chain was not
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secured in a position that required. unlocking the chain to remove it.
He had,
therefore, simply slipped the chain from over the top of the handwheel' and
opened the door.
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The inspector discussed this incident with the shift supervisor, and he
documented the event in CR 93-0143.
The nuclear equipment operator identified
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a way of locking the airlock doors. in a secure manner, and the doors were
properly locked. The shift supervisor instructed the firewatch ' personnel on-
the proper way of locking the airlock:and had an operator aid posted to remind
personnel operating the airlock on the proper way to lock it.
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On March 31, during 'another tour, the inspector noted an instance where the
outer door of the 171-foot elevation containment persennel airlock was not
.being locked as required by the Plant Manager's dircctive.
In this instance,
the person responsible for locking the door was using the padlock but, again,
the chain was draped over the handwheel in such a manner that it could be
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easily removed without unlocking the padlock. Whcn cha'llenged by the
inspector, the person explained that the chain was too short. After the
inspector called his attention to other chains available to him in the
airlock, the individual-switched chains around, thereby solving the problem.
.The inspector informed the shift supervisor of the continuing problem with
people not complying with Technical Specification 3.6.1.4.b, as directed by
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the Plant Manager.
CR 93-0160 was issued to document the second problem.
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Actions were taken to ensure that individuals responsible for locking the
. airlock doors clearly understood the Plant Manager's directive, and double
padlocks were utilized to capture the handwheels.
These issues were addressed
during an enforcement conference conducted at Region IV Headquarters on
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April 21, 1993, and were reviewed in terms of immediate corrective actions
taken in response to' containment breaches and inoperability of airlock
interlocks discussed in NRC-Inspection Report 50-458/93-11.
Therefore,.no
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further action is necessary under this inspection report.
During an NRC management tour of the plant conducted on March 25, 1993,
several deficiencies were'noted. A ventilation discharge screen was covered
with what appeared to be dead insects.
The licensee stated that the
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ventilation was nonsafety-related and that the condition had been identified
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on plant management's housekeeping punchlist prior to the-tour. Additionally,
the inspectors ~ noted a foam-like substance floating on the surface of the
auxiliary building crescent area sump.
This sump was the suction point.of the
. safety related suppression pool pumpback system.
The licensee evaluated the
substance and convinced the inspector that it did not pose a safety concern
and did not affect the operability of the system.
Housekeeping had generally improved, but during a reactor building tour near
the end of the report period, the inspectors noted a number of flashlights,
poly bags, and small tools in the containment that had the potential for
falling into the suppression pool.
This was brought to the attention of
licensee management and was promptly corrected.
The inspectors have noted
fewer ladders improperly stored and less debris in the pl mt.
The inspectors noted continued improvements in plant preservation.
The
licensee had just completed painting of the 123 foot elevation
"T" tunnel.
This took extensive effort and made a large difference in the appearance of
the "T" tunnel, which was used as an entrance area for the turbine, radwaste,
and auxiliary buildings, as well as the auxiliary control room.
3.3 Conclusions
Control room operator performance and professionalism was found to be
improving, based on routine daily observations and by observation of the
shutdown that was implemented on April 17 and 18, 1993.
4 MONTHLY MAINTENANCE OBSERVATIONS (62703)
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The station maintenance activities addressed below were observed and
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documentation reviewed to ascertain that the activities were conducted in
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accordance with the licensee's approved maintenance programs, the Technical
Specifications, and NRC Regulations.
4.1
Division 11 Standby Diesel Generator (DG) Outage
)
On April 14, 1993, the inspector observed maintenance activities associated
with a planned Division II Standby DG outage.
Three mechanical maintenance
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activities and part of one instrument and controls activity were observed.
The inspector reviewed the work packages and clearances.
The prejob briefing
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was attended and work followed to completion.
The tasks were MWO P564471,
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" Lubricate Diesel Keepwarm Oil Pump, lEGO*PIB"; MWO R170898, " Valve Leaks by
Seat-Repair, lEGA*V137"; MWO R170881, " Inspect Division II EDG Air Start
Distribution per CR 93-0035."
All the MW0s were under Clearance RB-1-93-6288.
The work packages were well
prepared with clear work instructions and clearly defined inspection hold
points.
During the valve repairs, care was taken to prevent foreign material
intrusion into the open pipe.
The inspection of the air start system involved
a custom test fixture to ensure an adequate vent path from the DG cylinders
after air pressurization during the start cycle. The system engineer was
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present during the work activities to explain the intent of. the work and
monitor progress. The inspector also observed the quality control inspections
during the valve repair. The quality control inspector verified the condition
of the parts upon disassembly, foreign material exclusion practices, condition
of materials upon reassembly, including seat contact area by confirming blue
checks, proper qualification of the gasket material, and final torque values
upon reassembly.
All work observed was performed in accordance with the work
instructions provided.
The inspector also observed portions of MWO P562478, a calibration of the.
alarm for the diesel Jacket outlet water temperature. One deficiency was
noted with the package. The inspector observed that the calibration directed
the technicians to perform calculations to compensate 'for the ambient
temperature due to the dissimilar metal junction introduced by the test leads
in the alarm cabinet, but did not provide space for the technicians to record
actual uncompensated values found, ambient temperature, and the corresponding
millivolt value to be subtracted, or the values resulting from the
calculation. This resulted in loss of actual values measured by the equipment
prior to the subtraction of the compensation. The foreman could not review
the calculations for accuracy prior to signing the package, because they were
not recorded. This observation, to enhance the maintenance calibration
process, was discussed with the licensee.
4.2 Rebuild of a Recirculation Pump Seal Package
On April 19, 1993, the inspector observed portions of the work performed under
MWO R175003. This MWO was written to rebuild the spare Seal
Package IB33*PC001B for installation on Reactor Recirculation Pump B.
The
inspector verified that the replacement parts to be used were acceptable for
installation in safety-related applications in accordance with the requisition
ticket.
Seal package body components, seals, and o-rings were verified by
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serial number to ascertain that the appropriate part was being utilized.
The inspector observed that the repairmen were utilizing the work package job
plan and Corrective Maintenance Procedure CMP-9020, "B33*C001 Reactor Water
Recirc. Pump Disassembly, Inspection, Rework and Reassembly." The individuals
were verified to be qualified for the work in accordance with the licensee's
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training procedures.
The individuals appeared to be knowledgeable of the job
and professed that they had performed the procedure in the past.
The
repairmen meticulously took measurements of parts to be reused as well as
those to be replaced.
The inspector noted that there was full time coverage of the job by a plant
quality control inspector. The inspection report indicated that all
observations were satisfactory.
Acceptable tolerances had been measured on
all parts. Hold points were met throughout the observation. Additionally,
the licensee had requested and procured the services of representatives from.
the pump manufacturer to assist as technical advisors in the rebuild effort.
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The inspector observed the radiation work practices used during the effort.
'The rebuild was performed within a contaminated zone in the hot machine shop.
At one point.the quality control inspector noted that the craftsmen did not
have enough room to pass one of the radiological boundaries without moving it
to one side.
The radiation protection technician was notified and the zone
was enlarged to provide adequate room as well as appropriate contamination
control.
4.3 C;nclusions
The performance of maintenance observed during this inspection period was
good.
Procedures provided by the MW0s appeared to be appropriate in each
case.
5 BIMONTHLY SURVEILLANCE OBSERVATIONS (61726)
The inspectors observed the surveillance testing of safety-related systems and-
components addressed below to verify that the activities were being performed
in accordance with the licensee's approved programs and the Technical
Specifications.
5.1 Retest of Replaced Potter & Brumfield MDR Relays
On April 23, 1993, the inspector observed portions of Surveillance Test
Procedure STP-511-4802, Revision 3, " Main Steam Line Isolation-Main Steam Line-
Radiation High, 18 Month Response Time Test, (D17-K610B,C) Channels B and C."
This test was designated in MWO 166031 as a retest following replacement of
Potter & Brumfield MDR-4171 Relay IC71A*K78.
The technicians followed the procedure in a step-by-step manner, demonstrated
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good communications with the operators, and followed the administrative
requirements for lifted leads.
A quality control inspector was present to
observe the test.
The technicians experienced problems with the frequency generator and noted
that voltage transients were causing partial trips.
Because they were unable
to resolve the problem by changing test equipment connecting wires, they
stopped the test and restored the system to normal. After. consulting with the
system engineer, the MWO was revised to require a much simpler test,
STP-511-4502, Revision 5, "RPS/ Isolation Actuation-MSLI-Main Steam Line
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Radiation-High Monthly Channel Functional (D17-K610B)."
Because of the difference in the two tests, the inspector questioned how the
simpler test could meet the retest requirements.
The foreman responded that
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they were able to take the bench test time response of the replacement relay
and add it conservatively to the previous system time response results from
the last refueling outage, leaving only the channel functional test to be
performed.
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In order to evaluate the adequacy of retesting specified for the replacement
.of MDR relays, the inspector requested the licensee to demonstrate in detail
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how the selected surveillance test procedures were verified to adequately
retest the new relays.
This was scheduled to occur on May 4, during the next
inspection period. The results will be documented in a future inspection
report under Inspection Followup Item 458/93010-2.
6 ENGINEERED SAFETY FEATURE WALKDOWN (71710)
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The objective of this inspection was to perform a detailed walkdown of a
representative sample of the accessible portions of the control room air
conditioning system to verify the system's capability to perform its intended
safety functions.
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6,1
System Status
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The control room air conditioning system, *;ith two independent air handling
unit / filter train subsystems were in an sperable status during power
operations and while shut down for Plan led Outage 93-01.
6.2 Procedure Reviews
The inspectors commenced a review of System Operating Procedure SOP-0058,
Revision 7, " Control Building HVAC System," for adequacy and technical
agreement with the system piping and instrument drawing.
The review was
incomplete as of the end of this inspection period and will continue into the
next inspection period and be documented in NRC Inspection
Report 50-458/93-19. No discrepancies were identified.
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ATTACHMENT
1 PERSONS CONTACTED
1.1 Licensee Personnel
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R. E. Barnes, Supervisor, Maintenace Engineering
- J. B. Blakeley, Assistant Plant Manager, System Engineering
- J. E. Booker, Manager, Nuclear Industry Relations
- J. W. Cook, Senior Technical Specialist
D. R. Derbonne, Assistant Plant Manager. Operations, Radwaste & Chemistry
L. L. Dietrich, Supervisor, Nuclear Licensing
R. G. Easlick, Radwaste Supervisor
- L. A. England, Director, Nucl.!ar Licensing
- A. O. Fredieu, Supervisor, Maintenance Services
P. E. Freehill, Assistant riant Manager - Outage Management
- K. D. Garner, Licensing Ergineer
- K. J. Giadrosich, Supervisor, Quality Engineering
E. L. Glass, Supervisor, Instrument & Control
- P. D. Graham, Vice President (RBNG)
- J. R. Hamilton, Manager-Engineering
- W. C. Hardy, Radiation Protection, Supervisor
- G. R. Kimmell, General Maintenance Supervisor
- G. D. Lipham, Chemistry Supervisor
- D. N. Lorfing, Supervisor, Nuclear Licensing
- l. M. Malik, Supervisor, Operations Quality Assurance
C. R. Maxson, Senior Compliance Analyst
J. Mead, Supervisor, Electrical and Special Projects
T. G. Murphy, Director, Management Systems
W. H. Odell, Director, Radiological Programs
R. L. Roberts, Electrical Maintenance Supervisor
- J. P. Schippert, Plant Manager
B. R. Smith, Mechanical Maintenance Supervisor
- M. A. Stein, Director, Design Engineering
- K. E. Suhrke, General Manager, Engineering and Administration
- W. J. Trudell, Assistant Operations Supervisor
R. J. Vachon, Senior Compliance Analyst
- J. E. Venable, Operations Supervisor
C. W. Walker, Supervisor, Operations Quality Control
S. L. Woody, Director, Nuclear Station Security
- Denotes personnel that attended the exit meeting.
In addition to the
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personnel listed above, the inspectors contacted other personnel during this
inspection period.
2 EXIT MEETING
An exit meeting was conducted on April 27, 1993.
During this meeting, the
inspectors reviewed the scope and findings of the report.
When the inspectors
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summarized the issue in Section 4.1 of the report, pertaining to a diesel
jacket outlet water temperature alarm calibration procedure not requiring the
technician to record actual uncompensated values found, the inspectors
emphasized the importance of recording raw data for future reference so that
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calculated values could-be verified.
Licensee management ackowledged the
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concern'and. stated that they would review the matter and advise the inspectors
of their. findings and corrective actions.
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The licensee did not_ identify as proprietary,.any information provided to, or.
reviewed by the inspectors.
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