ML20137G041
| ML20137G041 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 03/21/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20137F997 | List: |
| References | |
| 50-498-97-01, 50-498-97-1, 50-499-97-01, 50-499-97-1, NUDOCS 9704010295 | |
| Download: ML20137G041 (13) | |
See also: IR 05000498/1997001
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION IV
Docket Nos:
50-498, 50-499
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License Nos:
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Report No:
50-498/97-01, 50-499/97-01
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Licensee:
Houston Lighting & Power Company
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Facility:
South Texas Project Electric Generating Station,
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Units 1 and 2
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Location:
8 Miles West of Wadsworth on FM 521
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Wadsworth, Texas 77483
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Dates:
January 12, through February 22,1997
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Inspectors:
D. P. Loveless, Senior Resident inspector
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J. M. Keeton, Resident inspector
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W. C. Sifre, Resident inspector
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Approved by:
J. l. Tapia, Chief, Project Branch A
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Division of Reactor Projects
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9704010295 970321
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ADOCK 05000498
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EXECUTIVE SUMMARY
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South Texas Project, Units 1 and 2
NRC Inspection Report 50-498/97-01;50-499/97-01
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This resident insm,ction included aspects of licensee operations,~ engineering, maintenance,
and plant support. The report covers a 6-week period of resident inspection.
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Operations
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Operations management clearly delineated the expectations for component
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operation including management's goal of zero mispositioned components
(Section 01.1).
A licensed operator performed a reactivity manipulation prior to exiting the control
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room without performing a proper turnover of responsibilities (Section 01.1).
Operators' discharged their licensed duties during observed activities in both units in
a notable fashion (Section 01.1).
Safety system equipment material condition and alignment were found to be good
(Sections 02.1 and O2.2).
One violation was identified when a procedural omission resulted in the loss of
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reactor coolant system level sight glass indication during a vacuum fill evolution.
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This avent-identified violation was not considered for enforcement discretion in
accordance with Section Vll.B.1 of the NRC Enforcement Policy because of the
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regulatory significance of the controls governing reactor vessel water indication
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while conducting midloop operations, and because the corrective actions for
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Violation 498/96004-01 should have preverited this event (Section 03.1).
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Maintenance
Procedural guidelines and management expectations were not met during the repair
of a main steam system valve. Although no actual safety significance was
identified, the quality of the work activity was considered poor based on the number
of anomalies identified during the observation (Section M1.1).
Observed surveillance activities were performed in accordance with Technical
Specifications requirements with no discrepancies identified (Section M1.2l.
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Enaineerina
Coordination among the engineering, operations, and maintenance organizations
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was observed to be very good upon discovery of a problem during .aview of
Generic Letter 96-06 (Section E2.1).
Plant Sucoort
Two minor examples of poor radiological work practices were identified during
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nearly 100 observations of workers and worker practices (Section R1.1),
The radiological controls, maintenance of emergency response facilities and
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equipment, and physical security activities observed were appropriately conducted
and controlled (Sections R1.1, P2.1, P2.2, and S1),
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fhport Details
Summary of Plant Status
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Unit 1 began this inspection period at 100 percent power. On January 25, the unit was.
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removed from service for the purpose of testing rod cluster control assemblies. Following
the completion of testing, Unit 1 was returned to full power on January 27, and remained
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at full power for the duration of the inspection period.
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Unit 2 began this inspection period at 100 percent power. On February 6, reactor power
was reduced to begin Refueling and Equipment Outage 2REOS. In addition to normal
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refueling activities, major work activities included main generator inspections and torque
adjustments and nondestructive examination of 100 percent of the steam generator tubes.
At the end of this inspection period, Unit 2 was in Mode 5.
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I. Operations
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Conduct of Operations
01.1 Control Room Observations (Units 1 and 2)
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Insoection Scoce (71707)
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. Using Inspection Procedure 71707, the inspectors routinely observed the conduct of
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operations in the Units 1 and 2 control rooms during normal full power operation,
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shutdown and startup operations, and refueling operations in Unit 2. The
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observations included: frequent reviews of control board and engineered safety
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features equipment status; routine attendance at shift turnover meetings;
observations of operstor performance; and reviews of control room logs and
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documentation. In addition to full power operation, the inspectors observed
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portions of the following evolutions-
Use of the boron thermal regeneration system (January 12 - 25)
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Unit 1 plant shutdown for rod drop testing (January 24)
Unit 1 reactor startup (January 26)
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Unit 2 plant shutdown for refueling outage (February 6 and 7)
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Unit 2 plant cooldown and reactor system venting to atmosphere
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(February 8)
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Unit 2 reactor coolant system drain to midloop for nozzle dam installation
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(February 11)
Unit 2 refueling activities (/ebruary 14 and 15)
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Unit 2 reactor coolant system drain to midloop for nozzle dam removal
(February 19)
Unit 2 reactor coolant system vacuum fill (February 20)
b.
Observations and Findinas
The inspectars observed good procedure usage in the control rooms.
Communications were generally formal. Annunciator alarm responses were prompt
and the alarm status appropriately dispositioned. -Operators' use of self-verification
techniques was evident. The engineered safety features systems in both units were
verified to be aligned in accordance with Technical Specifications requirements
during various plant operating conditions.
On January 24, the inspectors attended a discussion of component mispositioning
consequences given by the Unit 1 Operations Manager. The discussion clearly
stated management's expectations regarding component operation to the oncoming
operators from both units. Through interviews following the discussion, the
inspectors ascertained that operators were fully aware of their responsibilities and
the corrective actions being taken to achieve management's goal of zero
, components mispositioned.
On January 31,1997, the inspectors observed the primary reactor operator perform
a dilution of the reactor coolant system. The operator added 20 gallons of
deionized water to the volume control tank using the automatic setting on the
integrator. Self-verification of handswitch manipulations was evident.
The inspectors noted that the primary operatoi had not discussed the dilution with
any of the control room personnel, in addition,4 minutes after completing the
evolution, the operator left the control room without informing the secondary
operator.
The inspectors discussed this with the shift supervisor. He stated that the primary
operator's performance clearly did not meet expectations and that the issue would
be discussed with the operator.
c.
Conclusions
The inspectors concluded that, in general, the operators continued to perform their
duties in a professional manner with an observed exception. On one occasion, a
licensed operator performed a reactivity manipulation and then exited the control
room without properly briefing his relief.
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02
Operational Status of Facilities and Equipment
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02.1 Plant Tours (Units 1 and 2)
a.
Insoection Scoce (71707)
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The inspectors routinely toured the accessible portions of plant areas in Units 1 and
2. Areas of special attention during this inspection period included.
Units 1 and 2 isolation valve cubicles.
Units 1 and 2 mechanical auxiliary buildings.
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Unit 2 fuel handling building.
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Unit 2 turbine generator building.
Unit 2 reactor containment building.
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b.
Observations and Findinas
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The inspectors observed that in both units, systems and components were being
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maintained in good material condition. However, on several occasions, minor fluid
leakage from various plant components was identified and reported to control room
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personnel. Particular ettention to maintaining internal systems cleanliness was
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observed during the Unit 2 refueling outage. Open systems and components were
appropriately covered to prevent entry of foreign debris. Good attention to detailin
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securing portable gas cylinders was observed. Plant chemistry was appropriately
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maintained for the various plant conditions in both units.
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The inspectors toured the Unit 2 reactor containment building during the refueling
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outage. The inspectors visually inspected reactor coolant system, chemical and
volume control system, and emergency core cooling system valves and piping inside
the bioshield that were normally inaccessible. Overall conditions were good.
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Conclusions
The inspectors concluded that the material condition of systems and components
observed in both units was good,
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02.2 System Flowcath Alianments (71707)
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On February 4, the inspector verified the flow path alignments of the auxiliary
feedwater pump discharge lines in the isolation valve cubicles. All four lines in both
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units were inspected. No valve alignment discrepancies were identified. However,
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two feedwater preheater lines in Unit 2 were mislabeled. In addition, the inspectors
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verified that the fused disconnects associated with the steam-driven pump steam
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admission valves in both units were closed.
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On January 18, the inspectors verified the flow path alignment of the spent fuel
pool cooling and cleanup system filters No discrepancies were identified.
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Operations Procedures and Documentation
O3.1 Loss of RCS Sicht Glass Durino Vacuum Fill
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Inspection Scope (71707)
On February 20, the inspectors observed the Unit 2 control room staff performing a
fill of the reactor coolant system under vacuum from a reduced inventory condition.
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b.
Observations and Findinas
The licensed operators performed the fill of the reactor coolant system in
accordance with Plant Operating Procedure.0 POP 03-RC-0100, Revision 4, "RCS
Vacuum Fill." Approximately 10 minutes after starting the vacuum fill process, a
plant operator reported the loss of levelindication on the reactor coolant system
magnetic sight glass. The sight glass was one of five independent levelindications
required by Plant Operating Procedure OPOP03-ZG-0009, Revision 16, "Mid-Loop
Operation." Both Procedures OPOP03-RC-0100 and OPOP03 ZG-0009 required the
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magnetic sight glass to be in service. The shift supervisor imraediately suspended .
the vacuum fill, contacted the system engineer, and c'irected reactor operators to
walk down the sight glass connections to check for air inleakage.
The sight glass indication returned to normal when the reactor coolant system
pressure was returned to atmospheric. During the system walk down, a reactor
operator identified an open vent valve connected to a temporarily installed vent
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manifold used only during outages. The open valve permitted air to flow into the
sight glass, forcing the coolant levelin the sight glass to drop to an undetectable
level during the vacuum fill. Following closure of the valve and verification of '
system alignment, the vacuum fi.ll was resumed and completed with no further
perturbations. Condition Report 97-3721 was written to address and correct the
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alignment discrepancy.
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The inspectors determined, through interviews with the midloop coordinator, the
shift supervisor, and the system engineer, that the identified vent valve had been
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previously opened in accordance with Plant Operating Procedure OPOP03-ZG-0007,
Revision 13, " Plant Cool Down." The Procedure OPOP03-ZG-0007 did not instruct
the operators to close the valve and the valve was not included in either the
vacuum fill procedure alignment checklist nor the midloop operation valve
alignment. The inspectors reviewed NRC inspection Report 50-498/96-04;
50-499/96-04 which included Violation 498/96004-04 th7t addressed a situation
wherein both nnrrow range reactor coolant system levelinstruments were unable to
perform their required function because of misaligned valves. These valves were
also not included in the midloop operations procedure valve alignment.
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Conclusions
The inspectors concluded that, contrary to the stated procedura. requirements, the
open valve on the upper connection assembly during the vacuum fill evolution
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rendered the magnetic sight glass unable to perform its function. This was
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identified as Violation 499/97001-01. This event-identified violation was not
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considered for enforcement discretion in accordance with Section Vll.B.1 of the
NRC Enforcement Policy because of the regulatory significance of the controls
governing reactor vessel water indication while conducting midloop operations, and
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because the corrective actions for Violation 498/96004-01 should have prevented
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this event.
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II. Maintenance
M1
Conduct of Maintenance
M1.1 General Comments on Field Maintenance Activities
a.
Insoection Scoce (62707)
~ The inspectors observed portions of the following work activities identified by their
work authorization numbers:
Unit 1:.
101931
Steam Generator 18 Power Operated Relief Valve Stroke Time
Exceeded Surveillance Test Requirements
Unit 2:
101148
Replace Steam Generator 2A Level. Feedwater Flow, Steam
Flow Recorder in the Unit 2 Control Room
103029
Pressurizer Pilot-Operated Relief Valve failed to meet its
Required Stroke Time during Surveillance Testing
b.
Observations and Findinas
The inspectors found that the work performed during two of these activities was
conducted in a thorough and professional manner. The work was performed by
knowledgeable, qualified technicians utilizing approved procedures. Supervisors
were observed providing an appropriate level of oversight. System engineers were
observed providing quality technical support as needed. Prejob briefings were
thorough and radiological controls were in place where applicable. However,
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exceptions to these observations were noted during the repair activities associated
with a main steam system valve.
On January 24, the inspectors observed repairs performed on Steam Generator
Power Operated Relief Valve 1B. This work was being conducted as authorized by
Work Order 356332, " Steam Generator 1B PORV Stroke Time Exceeded
Requirements of Plant Surveillance Procedure OPSP03-MS-0001." During the
replacement of the valve actuator hydraulic accumulator, the inspector observed
some poor practices during the performance of Plant Maintenance
Procedure OPMPO4-SG-0007, Revision 6, " Steam Generator PORV Hydraulic
Actuator Maintenance."
The parts removed from the actuator were laying in several locations and none of
them were labeled, tagged, or placed in labeled or marked containers at time of
disassembly to ensure accountability and traceability of parts during maintenance,
in addition, contrary to the expectation that procedure performers enter a check
mark in the corresponding blank when each procedure step is completed, craftsmen
quickly checked off a number of steps that had previously been ccmpleted upon the
inspectors' arrival.
. Step 6.15.6 of Procedure OPMPO4-SG-0007 directed the performer to charge the
accumulator with nitrogen to a predetermined pressure. During this evolution, the
pressure gauge responded so quickly to the pressure increase that the craftsmen
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had to open a line fitting to release pressure and bring it back to the proper
accumulator pressure. The inspectors questioned whether the accumulator had
actually been pressurized or whether only the test line had been. The craftsmen
stated that the accumulator wai, properly charged and checked off the step in the
procedure. Following further questioning, the work supervisor directed the
craftsmen to reconnect the test rig and verify the accumulator pressure. The
pressure check indicated that the accumulator had not been charged. The
craftsmen determined that the accumulator fill valve on the accumulator had not
been opened during the original attempt at charging the system. The accumulator
was then properly charged and verified.
After work completion, craftsmen attempted to cycle the valve. The inspectors
made the following observations during the performance of this step: (1) the
craftsmen did not appear to utilize self-checking methods during the manipulation of
the valves; (2) the craftsmen first positioned both solenoids then utilized the
actuator hydraulic drain valve to cycle the operator; and (3) following the first
attempt to cycle the valve, the actuator did not stroke. The work supervisor
checked the valve alignment and determined that one of the solenoids had not been
properly positioned. The problem was corrected and the valve then stroked open.
Although no actual negative safety consequence resulted from the observed poor
practices, procedural guidelines and management expectations were not met during
this evolution. All work was performed with the valve manually isolated from the
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main steam system. Additionally, the scheduled postmaintenance testing would
have identified an uncharged accumulator or a mispositioned solenoid valve. The
Unit 1 Maintenance Manager agreed that the observed poor practices did not meet
management's expectations.
c.
Conclusions
Procedural guidelines and management expectations were not met during the repair
of a main steam system valve. Although no actual safety significance was
identified, the quality of the work activity was considered poor based on the number
of anomalies identified during the observation. The other activities observed were
conducted in a professional manner and personnel involved were thorough and met
management's expectations for the implementation of the maintenance program.
M1.2 General Comments on Surveillance Testina
a.
inspection Scoce (61726)
The inspectors observed portions of the following surveillance activities:
, Unit 1:
Plant Surveillance Procedure OPSP10-DM-0003, Revision 4, " Automatic
Multiple Rod Drop Time Measurement."
Unit 2i
Plant Surveillance Procedure OPSP06-RC-0003, Revision 4, "Undervoltage
Reactor Coolant Pump Relay Channel Calibration and Trip Actuation Device
Operational Test."
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Plant Surveillance Procedure OPSP06-RC-0004, Revision 4, "Under frequency
Reactor Coolant Pump Relay Channel Calibration and Trip Actuation Device
Operational Test."
Plant Surveillance Procedure OPSP10-DM-0003, Revision 4, " Automatic
Multiple Rod Drop Time Measurement."
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Observations and Findinas
The inspectors found that the observed surveillance activities were performed in
accordance with approved procedures. The surveillance procedures properly
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implemented the associated Technical Specifications requirements. The test
instruments utilized were within current calibration cycles. Expected annunciator
alarms were communicated to the control room operators prior to actuation.
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On January 25, Unit 1 operators performed a reactor shutdown to facilitate rod
cluster control assembly drop time testing. The inspectors observed testing
conducted in Unit 2 on February 8 during Refueling and Equipment Outage 2RE05.
The purpose of the surveillance activities was to obtain data for further evaluation
of the incomplete rod insertion problems experienced by both units during the last
few shutdowns. This data was reviewed by the NRC Office of Nuclear Reactor
Regulation personnel.
c.
Conclusions
Observed surveillance activities were performed in accordance with Technical
Specifications requirements with no observed discrepancies.
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Miscellaneous Maintenance issues (92902)
M8.1 (Closed) Unresolved item 498:499/96006-02: review the adequacy of intentionally
entering a Technical Specification Action Statement with no allowed outage time for
the purpose of performing corrective maintenance. This item was opened to
address two questions:
Was it appropriate to enter a Technical Specification action statement that
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had no allowed outage time in order to make repairs to plant components?
The inspectors noted that this issue had been previously addressed by the
, Technical Specifications Branch of the Office of Nuclear Reactor Regulation
as a generic concern when they answered the following question: "If the
[ Allowed Outage Time]is so short that repairs are impossible during that
time period, can the licensee remove a component from service with the
intention of completing the repair during the shutdown time, and risk a ',asty
plant shutdown if they subsequently find the repair cannot be compleud in
that time?" in the response the NRC determined that circumstances may
arise when plant safety is better served by delaying a shutdown action to
provide a safer configuration for a shutdown transient or to avoid an
unnecessary shutdown transient. NRC guidance still recommends not
entering a Technical Specification action statement unless the work can be
completed and the system returned to service within the allowed outage
time. However, if a licensee responsibly concludes that plant shutdown
should be delayed or corrective action can be accomplished so that an
unnecessary plant transient can be avoided, such a decision is permitted as
long as the shutdown times specified by the Technical Specifications are
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ls it appropriate to perform quality assurance reviews of critical system work
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packages by telephone?
In responding to this question, licensee representatives provided evidence
that the decision made was sound and had been properly controlled by
quality assurance procedures. A documented position had been previously
taken that inspections for soldering on nonmodification work was not
necessary. The basis for this decision was well documented and met the
criteria delineated in Plant Quality Procedure OPOP01-ZA-0005, Revision 5,
" Quality inspection Planning." In addition, independent inspection of the
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activity was provided on occasion and unscheduled inspections in this area
had indicated a 100 percent acceptance rate.
Ill. Enaineerina
E2
Engineering Support of Facilities and Equipment
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E2.1
Review of a Modification to the Reactor Coolant Drain Tank Containment
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Insoection Scone (37551)
The inspector reviewed the design control process, the actual design.of the
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modification, and implementation of the modification.
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b.
Observations and Findinas
During development of a response to Generic Letter 96-06, licensee engineeis
performed an analysis of an uninsulated containment penetration used to drain the
reactor coolant drain tank to the liquid waste processing system. A potential
overpressurization condition during a design basis accident was identified that could
have resulted in piping stresses exceeding the American Society of Mechanical
Engineers Code limits. Additional analysis indicated that piping stress limits would
not be exceeded if the penetration were insulated. The inspector reviewed
Condition Report Engineering Evaluation 96-12151-19, Design Change
Package 97-1064-2, and the response to Generic Letter 96-06 (ST-HL-AE-5554).
Coordination among the engineering, operations, and maintenance organizations
was observed to be very good. Engineering personnel provided prompt feedback to
control room operators once the extent of the problem was determined. Licensed
operators identified an alternate drain path from the reactor coolant drain tank and
placed a contingency procedure in effect until the modification could be
implemented. On January 23, the inspectors observed control room operators drain
the reactor coolant drain tank through this alternate drain path. A written plan of
action had been developed and was properly implemented. The drain line
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penetration was isolated and drained until the insulation could be installed.
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The drain line penetrations were returned to service and the normal drain paths
were reestablished following the insulation installation. Documentation for the
design change was in accordance with the licensee's requirements. The package
included appropriate reviews and approvals. The calculations for the addition of the
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insulation appeared to be proper. An appropriate 10 CFR Section 50.59 screening
was included.
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Conclusions
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.The engineering department's response to-this containment penetration design
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problem was performed in a timely fashion. It was well coordinated with control
room operators and maintenance personnel. The corrective actions and design
change were well documented.
E3
Engineering Procedures and Documentation
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E3.1
Safety Iniection System Loaic Circuit Not Fully Tested by Surveillance Procedure
, During a review of surveillance procedures in response to Generic Letter 96-01,
licensee engineers discovered that a quarterly actuation logic test for containment
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sump switchover did not require testing of the entire interlock logic circuit. Upon
discovery of the condition, operators in both units entered the action statement of
Technical Specification 3.0.3. This allowed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to correct the problem.
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However, Technical Specification 4.0.3 permitted an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to
complete the required testing prior to completing the actions of Specification 3.0.3.
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The circuits in both units were appropriately tested. Both units equipment tested
satisfactorily and the action statements were exited within the 24-hour time
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allowed. This issue will be further reviewed upon issuance of the licensee event
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report in a future inspection.
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E3.2 Review of Uodated Final Safety Analysis Report (UFSAR) Commitments
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A recent discovery of.a licensee operating their facility in a manner contrary to the
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UFSAR description highlighted the need for a special focused review that compares
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plant practices, procedures, and/or parameters to the UFSAR descriptions. While
performing the inspections discussed in this report, the inspectors reviewed the
applicable portions of the UFSAR that related to the areas inspected. The inspectors
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verified that the UFSAR wording was consistent with the observed plant practices,
procedures, and/or parameters.
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IV. Plant Support
R1
Radiological Protection and Chemistry Controls
R1.1 Tours of Radioloaical Controlled Areas
a.
Insoection Scoce (71750)
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On a routine basis, inspectors toured the accessible portions of the radiological
controlled areas in both units. During the Unit 2 outage, the reactor containment
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building was toured several times and plant workers were observed performing
radiation protection activities.
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b.
Observations and Controls
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Doors and gates inside the Unit 2 containment that were required to be locked for
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the purpose of radiation protection were found properly secured. Labels and
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postirigs established for the outage were in compliance with the licensee's
procedures. One minor example of poor contamination control practices and one
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, minor example of inadequate As-Low-As-is-Reasonably Achievable program
implementation were observed, reported to health physics personnel, and corrected.
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These two examples were considered isolated examples based on nearly 100
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observations of workers and worker practices.
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c.
Conclusions
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in general, radiological protection practices during normal operations and the Unit 2
outage were good. However, two minor examples of poor practices were noted.
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R1.2 Secondarv Chemistry Controls
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The inspectors routinely reviewed secondary water chemistry reports and radiation
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monitor alarm status. Secondary chemical analysis, the calculated primary to
secondary leak rate, and indication from the Nitrogen-16 radiation monitors all
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confirmed steam generator tube integrity. Review of the chemical analysis results
provided evidence of management attention and commitment to maintaining
chemistry parameters within appropriate limits.
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R6
Radiological Protection & Chemistry Organization and Administration
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R6.1 Postina of Notices to Workers (71750)
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During routine tours, the inspectors observed the licensee's regulatory information
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bulletin boards. All notices to workers were posted in accordance with
10 CFR Section 19.11.
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P2
Status of Emergency Preparedness Facilities, Equipment, and Resources
P2.1
Emeraency Response Facilities (71750)
During area tours, the inspectors observed that the Technical Support Centers and
Operations Support Centers in both units were readily available and maintained for
emergency operatiort.
P2.2 Meteoroloaical Towers (71750)
The inspectors routinely observed indicatior of meteorological conditions in the
main control rooms of both units. The data obtained indicated that both the
10-meter and the 60-meter towers remained operable.
S1
Conduct of Security and Safeguards Activities
S1.1 Daily Physical Security Activity Observations (71750)
a.
Insoection Scope (71750)
, On a daily basis, the inspectors observed the practices of security force personnel
and the condition of security equipment.
b.
Observations and Findinas
Protected and vital area barriers were in good condition. Personnel access
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measures and equipment searches for contraband were observed on a daily basis.
On one occasion, temporary office trailers were brought into the protected area.
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Security officers properly searched each trailer upon entrance and skirting was
installed to prevent access underneath,
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c.
Conclusions
Daily security force activities were conducted in an appropriate manner. On one
occasien, controls over temporary office trailers entering the protected area were
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found to be good.
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ATTACHMENT
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
T. Cloninger, Vice President,. Nuclear Engineering
B. Dowdy, Manager, Operations, Unit 2
J. Groth, Vice President Nuclear Generation
S. Head, Licensing Supervisor
K. House, Supervisor, Design Engineering Department
B. Logan, Manager, Health Physics
R. Lovell, Manager, Operations, Unit 1
B. Masse, Plant Manager, Unit 2
G. Parkey, Plant Manager, Unit 1
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T. Waddell, Manager, Maintenance, Unit 1
INSPECTION PROCEDURES USED
IP 37551:
Onsite Engineering
IP 61726:
Surveillance Observations
IP 62707:
Maintenance Observation
IP 71707:
Plant Operations
IP 71750:
Plant Support
IP 92902:
Followup - Maintenance
ITEMS OPENED. CLOSED, AND DISCUSSED
Ooened
499/97001-01
Reactor coolant system level sight glass not in
service during reduced reactor coolcnt system
inventory operations.
Closed
498;499/96006-02
Review the adequacy of entering an action
statement with no allowed outage time for
corrective maintenance
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