ML18065B128
| ML18065B128 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/19/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18065B126 | List: |
| References | |
| 50-255-97-11, NUDOCS 9712300164 | |
| Download: ML18065B128 (25) | |
See also: IR 05000255/1997011
Text
U.S. NUCLEAR REGULA TORY COMMISSION
REGION Ill
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspectors:
Approved by:
97i2300164 971219
.
ADOCK 05000255
Q
50-255
50-255/97011 (DRP)
Consumers Power Company
212 West Michigan Avenue
Jackson, Ml 49201
Palisades Nuclear Generating Plant
27780 Blue Star Memorial Highway
. Covert, Ml 49043-9530
August 28 - October 17, 1997
M. Parker, Senior Resident Inspector
P. Prescott, Resident Inspector
Bruce L. Burgess, Chief
Reactor Projects Branch 6
EXECUTIVE SUMMARY
Palisades Nuclear Generating Plant
NRC Inspection Report No. 50-255/97011 (DRP)
This inspection reviewed aspects of licensee operations, maintenance, engineering and plant
support. The report covers a 7-week period of resident inspection.
Operations
The inspectors noted that operators were thoroughly prepared for a plant downpower and
main turbine valve testing evolutions. Reactor engineering, system engineering and the
procedure sponsor provided good support for these evolutions (Section 01.2).
Operators failed to ensure that service water system valves were closed, which could
have resulted in the potential draining of the component cooling water system in an
Appendix R design bases fire. This resulted in the plant operating the facility outside the
design bases for 1 O days following discovery of the condition (Section 01.3).
The licensee conservatively decided to shut down the plant due to a relatively minor
increase in containment unidentified leakage. The inspectors noted that control room
operators performed well in bringing the plant to hot shutdown.
The inspectors concluded that the licensee provided good management oversight during
the reactor startup, including the approach to critical with a reactivity manager and reactor
engineering stationed onshift to augment shift coverage. Good conservative decision
making took place on several occasions, specificaljy: to return the plant to a hot
shutdown condition by inserting regulating rods during troubleshooting and repairs to
CROM 39, to insert all regulating rods when the ECP was not achieved with all control
rods out, and to conduct a PRC meeting prior to continuation of a plant startup following
the ECP discrepancy (Section 01.5).
Maintenance
The inspectors noted the operators were challenged by a number of emergent equipment
problems during the plant shutdown. This was indicative that the licensee continues to
struggle with plant material condition issues (Section M1 .1).
The inspectors concluded that the maintenance procedure for repair of the waste gas
surge tank was inadequate for the circumstances. The procedure allowed the waste gas
surge tank to be vented to the auxiliary building atmosphere by allowing the gagging of
relief valve, RV-1114, resulting in the contamination of five individuals during a routine
VCT gas sample. The use of the procedure should have caused operators to question
the potential for a breach of the waste gas surge tank discharge piping. Also, adequate
equipment controls were not provided to prevent personnel contamination. The
inspectors concluded that the use of a fluted tap by maintenance personnel when a
2 inch threaded bolt was specified in the work procedure was inappropriate and
contributed to the contamination of personnel (Section M1 .2).
3
Engineering
The inspectors found the compensatory measures taken for the identified Appendix R
issues to be adequate. The Appendix R enhancement review was found to be
progressing slowly. However, the review appeared to be thorough (Section E1 .1).
Plant Support
The licensee's actions to improve the resin transfer process resulted. in an error-free
evolution for the spent fuel pool job (Section R1 .1).
4
~ .
Report Details
Summary of Plant Status
The plant operated at essentially full power for the inspection period until September 19, 1997. At
9:21 p.m., EST, a power reduction was commenced to 84 percent to perform turbine valve
testing and repacking of a heater drain pump. Operations returned the plant to full power on
September 21, 1997 at 8:00 a.m. On September 29, 1997, at 8:30 p.m., a plant shutdown was
initiated to facilitate repairs on a small leak on a primary coolant pump leakoff line. The turbine
was taken offline at 5: 14 a.m. on September 30, 1997. The reactor was subcritical at 11 :00 a.m.
The forced maintenance outage was completed on October 15, 1997, when the reactor went
critical at 1 :OO p.m. The generator was synchronized and breaker closed at 11 :39 p.m.
I. Operations
01
Conduct of Operations
01.1
General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing
plant operations. The conduct of operations was considered by the inspectors to be
good; specific events and noteworthy observations are detailed below.
01.2
Reactor Downpower and Main Turbine Testing
- a.
Inspection Scope (71707)
The inspectors observed the conduct of control room operations for the downpower to
repack heater drain pump P-108 and perform main turbine valve testing. Applicable
procedures were reviewed.
b.
Observation and Findings
On September 19, 1997, the inspectors observed control room operators commence a
downpower to 85 percent reactor power. The purpose of the downpower was to allow
operation with only one heater drain pump. Heater drain pump P-108 required repacking
due to excessive leakage and testing of the main turbine governor and stop valves was to
be performed. No operator performance weakness were noted. An extra nuclear control
operator was added to support the shift and the inspectors noted good operator
attentiveness to panels. A reactor engineering supervisor observed the downpower and
issued appropriate guidance to maintain a proper axial shape index curve. The
procedure sponsor was also present to monitor the downpower activity and verify the
adequacy of the downpower procedure. System engineering monitored vibration due to
concerns with the increased main turbine and generator vibrations caused by the missing
piece of shroud on the low pressure tui'bine rotor stage. No problems with vibrations
were noted.
Prior to testing of the main turbine governor and stop valves the operations shift had "just
in time" training on the simulator. A question was asked by operators regarding how to
5
back out of the surveillance should a problem occur requiring a rapid downpower. The
question was resolved prior to commencing the test by assigning an extra nuclear control
operator to enhance control room panel monitoring. System engineering monitored main
turbine and generator vibrations during the testing. No testing problems were identified.
c.
Conclusions
The inspectors noted thorough preparedness by operations for the downpower and main
turbine valve testing evolutions. Reactor engineering, system engineering and the
procedures sponsor provided good support for these evolutions.
01.3
Inadequate Appendix R Compensatory Measures
a.
Inspection Scope (71707)
The inspectors reviewed the licensee's corrective actions taken in response to a
reportable 50. 72 involving a condition outside the design basis. This condition is the
result of an Appendix R fire involving the component cooling water (CCW) and service
water (SW) seal cooling valves for the essential safety systems (ESS) pumps. The
- licensee identified that during a control room fire a hot short may fail open interfacing
CCW/SW systems valves resulting in the loss of all CCW water to the SW system. The
most limiting scenario could lead to draining of the CCW system in approximately
25 seconds.
b.
Observations and Findings
On September 12, 1997, as an interim compensatory measure for the potential Appendix
R component cooling water loss of inventory scenario, an auxiliary operator was directed
to place caution tags on the air supply valves to CV-0951, CV-0880 and CV-0879.
Caution tags were also placed on the respective control switches in the control room.
This compensatory measure was proposed as corrective action for condition
report C-PAL-97-1270, which had detailed the Appendix R scenario. The air supply
valves were required to be in the closed position to ensure the valves would not
inadvertently open.
The onshift shift supervisor and auxiliary operator located the three valves to be caution
tagged in the plant. The auxiliary operator manipulated the air isolation valve to ensure
they were open. This was observed by the shift supervisor. At this point, the shift
supervisor was uncertain how much the auxiliary operator comprehended about the
required task. Also, the auxiliary operator did not realize that to immediately resolve the
issue the desired position for the air isolation valves was closed. The onshift senior
reactor operator guidance to the auxiliary operator was to hang the tags. The auxiliary
operator was not specifically directed by the shift supervisor to place the valves in the
closed position. There is no procedural requirement that when caution tags are used,
plant equipment is verified to ensure it is left in the required position.
A nuclear control operator was tasked to hang the caution tags. The nuclear control
operator directed the auxiliary operator to hang the caution tags on the control valve air
isolation valves. At this point, the nucl.ear control operator was uncertain if he directed
the auxiliary operator to close and tag, or just tag the valves. The nuclear control
6
operator assumed the auxiliary operator understood the issue and knew what actions
were required to proper1y implement the caution tag requirements. The assumption was
based on the fact the auxiliary operator had ear1ier walked down the valves with the shift
supervisor.
The auxiliary operator proceeded to hang the caution tags on the air isolation valves. The
auxiliary operator did not close the air isolation valves because he understood that they
needed to remain open to maintain the control valves closed. This was reinforced by the
fact that during the walk down the auxiliary operator manipulated the valves in front of the
shift supervisor to show the valves were in the open position. The auxiliary operator does
not recall ever being told by either the shift supervisor or nuclear control operator that the
valves needed to be closed.
Guidance under the "Special Instructions "portion of the caution tags read, "Do not open
without SS permission." There was no specific direction to suggest to the operator that
the air supply valves were to be shut at the time the tags were hung.
On September 24, 1997, permanent placards were placed in the control room to indicate
that the air supply valves were permanently closed. The system checklist procedure was
revised with the normal position of the valves indicated as "closed." A different auxiliary
operator removed the temporary caution tags and attached permanent caution tags. The
auxiliary operator found the air supply valves were open and not closed.
The failure to ensure adequate compensatory measures were taken to address the
Appendix R concern is considered a Violation of 10 CFR 50 Appendix B, Criterion XVI,
- eorrective Action.* However, the inspectors reviewed this licensee's actions for this
self-identified item and determined this was a Non-Cited Violation consistent with
Section Vll.8.1 of the Enforcement Policy {NCV No. 50-255/97011-02).
c.
Conclusions
The onshift operations personnel failed to take adequate measures to ensure the air
supply valves to three SW valves were left in the proper valve configuration. Failure to
ensure the air supply valves were closed could have resulted in the potential draining of
the CCW system in the event of an Appendix R fire. This resulted in the plant operating
the facility outside the design bases for 10 days following discovery of the condition. This
was considered a non-cited Violation.
01.4
Reactor shutdown for Forced Maintenance Outage
a.
Inspection Scope {71707)
The inspectors observed the pre-job brief, simulator "just in time" training and the plant
shutdown for a maintenance forced outage.
b.
Observations and Findings
The control room operators commenced an order1y shutdown of the reactor on
September 29, 1997. The shutdown was initiated due to increased unidentified primary
coolant system leakage. Primary coolant system leakage had risen from an average of
7
0.05 gpm to 0.199 gpm over the last four days. Inspection of the containment identified
the source of the leak as a cracked weld on a seal package controlled bleedoff line for the
P-50A primary coolant pump.
Operators held a pre-job brief and simulator "just in time" training prior to commencing the
shutdown. The pre-job brief was thorough. Roles and responsibilities were discussed
between the members of the operations shift. The operations superintendent provided
management oversight of these activities and subsequent shutdown. The inspectors
noted that the nuclear control operator (NCO) responsible for control rods and reactor
power and the other NCO responsible for turbine load reduction had not previously
performed a reactor shutdown. An extra NCO was assigned to the shift. This NCO was
assigned responsibility for maintaining proper feedwater flow and monitoring other
balance of plant equipment.
The simulator instructor discussed in detail a December 2, 1995, event following a turbine
trip. A high startup rate was observed while withdrawing control rods to maintain primary
coolant system temperature. The instructor stressed that review of the event noted the
NCO attempted to control T _ with control rods. However, at that point in time reactor
power was at approximately 10*1 percent power and the control rods had little or no effect
on T -* The instructor indicated that T _ should be controlled by decay heat removal
through the turbine bypass valve.
During the first simulator practice at taking the turbine and generator offline, simulator
parameters were difficult to control for the operators. Initial simulator conditions caused
feedwater oscillations and a 4 ° F difference in temperature between T rer and T -* The
inspectors noted a weakness in three-way communication with the shift. A subsequent
rerun on the simulator with more normal shutdown conditions noted improved shift
performance.
The off-going shift supervisor conducted a brief with the on-coming crew prior to the
conduct of the normal shift turnover. The oncoming crew assumed the shift with the plant
at approximately 85 percent reactor power. The crew commenced a turbine load
reduction at 24 percent an hour. The downpower proceeded in an orderly manner.
However, at approximately 23 percent reactor power, a problem arose during the transfer
of electrical loads from the station power to startup transformer. The G bus breaker
252-402 would not close. The G bus supplies power to one of the two cooling tower
pumps, P-398. The inspectors noted a momentary loss of command and control
because the shift supervisor and control room supervisor were focused in the effort to
reclose the breaker. The impact of the loss of one of the cooling tower pumps would be
relatively minor on condenser vacuum at this power level. At this time, the operators
were also contending with xenon buildup and its impact on reactor power. Further
attempts to close the breaker were unsuccessful. The shift supervisor and control room
supervisor re-focused on the plant shutdown. At approximately 18 percent power,
operators noted that the condensate pump recirc valve CV-0730 was not opening as
expected. The control room supervisor quickly anticipated plant conditions and the
actions required to address the problem. At approximately 7 percent power the turbine
was taken offline, the main feedwater pump was taken offline and auxiliary feedwater
lined up. The failure of CV-0730 required the condensate pumps to be shut off to prevent
damage to the pump due to low flow conditions. The main steam isolation valves were
closed, which meant loss of the bypass valve to control primary coolant system
8
temperature. The automatic dump valves opened to control primary coolal"!t system
temperature. The inspectors noted one other discrepancy. During the control room
supervisor's discussion of the sequence of events that would occur due to the failure of
CV-0730, the NCO controlling reactivity believed he would control temperature with
control rod movements. In actuality, the NCO's function was to control reactor power.
This had been reviewed during the simulator training.
c.
Conclusions
The licensee conservatively decided to shut down the plant due to a relatively minor
increase in containment sump level. The inspectors noted that control room operators
performed well in bringing the plant to hot shutdown. A momentary weakness in
cpmmand and control was noted when the shift supervisor and control room supervisor
were overly involved in attempts to close breaker 252-402 to maintain cooling tower pump
P-398 online.
01.5
Startup From Forced Outage
a.
Inspection Scope
The inspectors observed the initial and subsequent successful attempts for plant startup
after completing a forced maintenance outage. The main reason for the outage was to
repair a cracked weld of the pump seal package controlled bleedoff line for P-50A primary
coolant pump. "Just in time" simulator training and pre-job brief for the initial startup were
also observed.
b.
Observations and Findings
On October 8, 1997, operators commenced heatup of the primary coolant system. On
October 9, 1997, at 9:05 EST, the plant exited cold shutdown. On October 10, 1997, the
oncoming crew received "just in time" training and conducted a pre-job brief in
preparation for the approach to critical. The inspectors noted both the simulator training
and pre-job brief were well conducted.
However, the off-going shift received the primary coolant pump high/low alarm.
Operations determined the alarm was caused by the purification demineralizer inlet relief
valve RV-2013 lifting. The relief valve lifted when the third letdown orifice stop valve
opened, which caused an excessive differential pressure across the demineralizer.
Radiation protection subsequently notified operations of water leakage in the auxiliary
building. This was traced to the vent hole in the bonnet of RV-2013. A condition report
was initiated to determine why the relief valve lifted. A primary coolant system leak rate
was performed. Leakage had increased from .033 gpm to .3 gpm.
Another problem occurred during performance of procedure R0-21, "Control Rod Drive
System lnter1ocks." Control rod drive mechanism (CROM) 31 acted sluggish compared to
the other control rod drives. Operations discussed the issue with system engineering. A
determination was made to perform R0-22, "Control Rod Drive Drop Timing," for
CROM 31 to ensure there was no mechanical binding which could prevent the control rod
from being inserted into the reactor core. The licensee subsequently decided to place
the plant in cold shutdown and in order to facilitate repairs to RV-2013 and CROM 31.
9
Cold shutdown was reached on October 11, 1997. The rod drive for CROM 31 was
replaced.
On October 13, 1997, with all repairs completed to CROM 31 and RV-2013, the licensee
commenced a reactor heatup to hot shutdown. On October 14, 1997, with the plant on
the approach to critical, CROM 39 was found to exhibit sluggish movement with respect
to the remainder of rods in Group 4. CROM 39 was subsequently declared inoperable. In
order to facilitate troubleshooting and repairs, management directed the plant to be
placed in a hot shutdown condition, requiring all rods to be inserted into the reactor core.
Troubleshooting indicated that the motor's auxiliary contactors did not makeup proper1y.
A decision was made to replace the contactors while holding the plant in a hot shutdown
condition.
On October 15, 1997, with the relay contactor replaced, the plant proceeded on with the
approach to critical with no further difficulties with CROM 39. During the approach to
critical all regulating rods were withdrawn without achieving a critical condition. Initial
review of conditions by the inspectors determined that with all rods out, the estimated
critical position {ECP) was within the bounds of the uncertainty analysis window indicated
in technical specifications, although reactor engineering had predicted a critical rod
position on regulating rod Group 4. The licensee subsequently inserted all regulating
control rods and requested a resample of the primary coolant system boron.
Reanalysis of primary coolant system {PCS) boron concentration determined that the
PCS boron concentration was within the limits of the established ECP critical boron
concentration. The licensee determined that the ECP anomaly was due to
boron-10 depletion. The estimated ECP was recalculated to compensate for depleted
boron-10 concentration. A plant review committee {PRC) meeting was convened to
review the condition report on ECP anomaly and agreed with reactor engineering's
conclusions that the anomaly was due to boron-1 O depletion. The ECP was calculated to
be within the TS limits of 1 percent anomaly.
The new ECP was calculated with a lower boron concentration resulting in criticality with
Group 4 rods partially withdrawn. Criticality was subsequently achieved on October 15,
1997, within limits of the reestablished ECP. The turbine was synchronized to the grid
without incident. During the startup and attempts to achieve criticality, the inspectors
noted good command and control with appropriate conservative decision making. Three
way communication and use of procedures were noted by the inspectors. Good
management oversight was also provided during the startup including the approach to
criticality, with a reactivity manager and reactor engineering stationed on shift to augment
shift coverage.
c.
Conclusions
The inspectors concluded that the licensee provided good management oversight during
the reactor startup including the approach to criticality with a reactivity manager and
reactor engineering stationed onshift to augment shift coverage. Good conservative
decision making took place on several occasions, specifically: to return the plant to a hot
shutdown condition by inserting regulating rods during troubleshooting and repairs to
10
08
08.1
CROM 39, to insert all regulating rods when ECP was not achieved on an all rods out
condition, and to conduct a PRC meeting prior to continuation of a plant startup following
the ECP discrepancy.
Miscellaneous Operations Issues (92701 and 92901)
(Closed) Violation 50-255/94014-1A: Operations failed to ensure that the control rod
drive mechanisms (CROMs) were mechanically locked prior to inserting a reactor trip
signal, resulting in the CROM racks dropping into the reactor vessel upper guide
structure. On November 7, 1996, preparations were being made to remove the reactor
vessel head. Operations discussed the status of the uncoupling of the CROMs, in
support of reactor vessel head removal, with Refueling Services. It was understood that
the Refueling Services procedure was the controlling document. Operations then
- withdrew 44 of the 45 control rod drive racks. However, control rod drive (CRD)
number 33 was stuck. Maintenance personal began troubleshooting CRD 33.
Operations, in a subsequent shift turnover, failed to specify the Refueling Services
procedure as the document controlling the CRDMs. Maintenance, upon completing
repairs to CRD 33, withdrew it and mechanically locked CRO 33 in place. The shift
supervisor was notified of CRO 33 being repaired, withdrawn and locked into place. The
shift supervisor assumed the next step was to place the reactor protection system in the
reactor trip mode. The shift supervisor did not verify the status nor the controlling
procedure of the control rod drive racks. When the shift supervisor directed that the
reactor protection system be placed in the reactor trip mode, all control rod drive racks
except CRO 33 reinserted into the reactor. The Refuel Services procedure allowed the
control rod drive racks to be locked after all the racks were withdrawn .
On November 18, 1996, the licensee suspended all refueling work and conducted a
standdown meeting to reinforce nuclear, radiation and industrial safety concerns with all
work groups. Several events over the first two weeks of the outage were reviewed. A
common theme between events was the lack of communications between work groups
and alignment among workers.
Three specific responsibilities reinforced at the operations standdown meetings were:
Shift supervisors will identify operations activities from the outage schedule with
an understanding of the relationship between these activities and others. The
purpose of this action is to contribute to informed decision making within the
operations organization.
Work control senior reactor operators are to route work activities having
operations involvement to control room personnel for authorization. This action
will provide a direct exchange of information between work control and control
room personnel.
Control room personnel are to ensure they have a complete understanding of
activities.requested of them and that proper adjustments to work activities or plant
configuration have been made.
Also, the control rod drive blades and racks were inspected for damage due to the trip.
No damage was identified. This item is closed .
11
08.2
CClosedl Violation 50-255/9601+o1B: Operations shift did not return the isolation handle
Y-50 to the nonnal position prior to returning the bypass handle to the automatic position,
resulting in a loss of power to instrument AC bus Y-01. Circumstances that caused this
event were similar to those detailed in 50-255/96014-0lA. The event occurred on
November 17, 1996. Reasons for this event included inadequate understanding of the
work scope, inadequate communications, inadequate work control documents and
improper equipment operation.
The same specific responsibilities reinforced at the November 18, 1996 operations
department standdown meetings detailed in 50-255/96014-0lA above, were also part of
the corrective actions for this event. Additional corrective actions taken were:
All operations personnel involved discussed this event and the barriers that could
. have prevented it. The discussion included responsibilities for proper
communication, pre-job briefings, self checking and other aspects of operator
conduct.
The shift operations supervisor briefed all senior reactor operators on the need to
identify and conduct pre-job briefs. The expectation was reinforced to conduct a
pre-job brief whenever coordination between two or more work groups is required.
The maintenance and construction manager reinforced pre-job brief expectations
with maintenance and construction supervisors, using this event as an example.
This item is closed.
II. Maintenance
M1
Conduct of Maintenance
M1 .1 * General Comments
a.
Inspection Scope (62707 and 61726)
The inspectors observed all or portions of the following work activities:
Work Order No:
24713553:
27412530:
24514171:
.
24712252220:
081397HN01:
CV-0733 Slowdown isolation valve: Troubleshoot increased
P-558 charging pump: Repack and reassemble pump
P-558 Seal lubrication pump: Retenninate seal lube motor
Dry fuel storage cask: Loading, dry runs
Sluice T-50 Spent fuel pool demineralizer
12
24711013:
24711141:
Surveillance Activities
M0-38:
Heater drain pump P-108: Repack
Heater drain pump P-108: Fitting upstream of MV-HED114
leaking. Repair
Auxiliary Feedwater System Monthly Test Procedure
(P-8C)
SOP-8 Attachment 2: Testing of Main Turbine Valves/Protective Trips
GOP-12:
Heat Balance Calculation
b.
Observations and Findings .
The inspectors found the work performed to be professional and thorough. All work
observed was done with the work pack.age present and in active use. Work pack.ages
were comprehensive for the task and post maintenance testing requirements were
adequate. The inspectors frequently observed supervisors and system engineers
monitoring work. When applicable, work was done with the appropriate radiation control
measures in place.
c.
Conclusions
Overall, the inspectors observed good procedure adherence, maintenance and radiation
worker practices. Specific observations are detailed below.
M1.2
Volume Control Tank Gas Sample Leak into the Auxiliary Building
a.
Inspection Scope
On August 12, 1997, during a routine volume control tank (VCT) gas sample by a
chemistry technician, radioactive gas was released into the auxiliary building via the
waste gas system contaminating several individuals. The inspectors observed the
licensee's actions to identify the source of the system leakage and the impact on the
contaminated individuals.
b.
Observations and Findings
On August 12, 1997, waste gas compressor, C-50A, was tagged out of service (OOS)
due to ongoing maintenance. During the time the waste gas compressor was OSS, a
chemistry technician received permission from the operating shift to sample the VCT.
During the sampling process, the purged gasses from the VCT were discharged to the
waste gas surge tank room, subsequently contaminating the maintenance crew.
In reviewing the switching and tagging orders for the maintenance activity, the inspectors
determined that the VCT sampling should not have had any effect on the maintenance
activity. However, in reviewing the work instructions, Permanent Maintenance Procedure
WGS-M-2, "Inspection and Repair of Waste Gas Compressors, C-50A and C50B," the
13
inspectors noted that the procedure requires the use of a relief valve gagging device to
be installed on the discharge of C-50A. The procedure requires the use of a two inch
long bolt to be installed as a gagging device on relief valve, RV-1114. This gagging
device was installed to prevent an inadvertent relief valve lift during hydrostatic pressure
testing following repairs. However, a proper size bolt was not available and a fluted tap
was installed in its place. The installation of a gagging device resulted in breaching the
system boundary. The fluted tap verses a threaded bolt further compounded the situation
in that it resulted in a larger opening in the discharge piping. The installation of the
gagging device resulted in inadvertently venting the waste gas surge tank to atmosphere,
as the relief valve discharges to the waste gas surge tank. Thus once the VCT gasses
were purged to the waste gas surge tank, they were vented back through the relief
valve's discharge line to the auxiliary building atmosphere. The VCT sampling resulted in
a release of radioactive gases to the waste gas surge tank room where the maintenance
crew was working on the waste gas compressor. *All five individuals working on the
compressor at the time of the radioactive release were found contaminated. The failure
to adequately control the breaching of the relief valve discharge piping is considered a
violation of 10 CFR 50 Appendix B, Criterion V, in that Permanent Maintenance
Procedure WGS-M-2, was inadequate for the circumstances and resulted in the
contamination of the maintenance crew.
The operating crew was alerted to the high airborne conditions when radwaste area
monitor, RE-1809, alarmed and tripped the auxiliary building supply fan, V-10. Radiation
protection personnel were immediately notified and restricted access to the auxiliary
building and obtained air samples of the area. The released gases were subsequently
discharged to the stack and had an activity level of approximately 1300 counts per
minute. The radioactive release alert alarm setpoint was 1,300,000 counts per minute.
The maintenance personnel working in the area of the waste gas surge tank room were
informed of the problem and proceeded to access control. The repairmen were all
monitored for contamination and all five alanned the PCM-1 B monitors at access control.
All cleared the PM-7 monitors on egress to the protected area for whole body counting,
after being detained at access control to allow the activity to decay. The maintenance
workers were subsequently whole body counted with no positive results prior to leaving
the site.
c.
Conclusions
The inspectors conciuded that the maintenance procedure was inadequate for the
circumstances. The procedure allowed the waste gas surge tank to be vented to the
auxiliary building atmosphere by allowing the gagging of relief valve, RV-1114, resulting in
the contamination of five individuals during a routine VCT gas sample. The procedure did
not appropriately alert operations personnel to the potential for a breach of the discharge
piping. Therefore, adequate equipment controls were not provided to prevent personnel
contamination. The inspectors also concluded that the use of a fluted tap by
maintenance personnel when a two inch threaded bolt was specified in the work
procedure, was inappropriate and contributed to the contamination of personnel.
14
M1 .3
Complicating Factors on Plant Shutdown
a.
Inspection Scope (62707)
The inspectors observed the plant shutdown for the forced maintenance outage. Several
emergent equipment problems were noted.
b.
Observations and Findings
The main purpose of the outage was to perform a weld repair to a cracked section of
piping on the pump seal package controlled bleedoff line for P-50A primary coolant pump.
The P-50A pump had small seal flow oscillations and occasional seal pressure spikes~
The root cause for the line developing cracks had not yet been determined.
The original scope of the outage was significantly increased however due to several other
equipment problems encountered during the shutdown. Condensate recirc valve
CV-0730 failed to open at lower turbine loads. This forced operators to stop the
condensate pumps and complicated the shutdown. An air leak was found in the air
controller pneumatic relay assembly. The licensee determined to bring the plant to cold
shutdown instead of hot shutdown due to the emergent equipment problems. With the
plant in cold shutdown, the licensee also chose to replace the primary coolant
pump P-508 seal package. The upper stage of the seal package had destaged and the
other three stages compensate for the failed stage. The P-50A and P-50C also have seal
flow oscillations of .01 gpm and .05 gpm, respectively. Also, instrumentation cables for
the P-50A motor temperature indication had to be replaced due to damaged conduit
which allowed water inleakage from the cracked controlled bleedoff line. Failure of the
instrumentation was the means by which the licensee determined the suspected source
of the primary coolant system leak.
The shutdown was also complicated by the failure of the 1-G bus breaker 252-402 to
transfer from the station power to startup transformer. This forced the operators to shut
down cooling tower pump P-398. Subsequently, the breaker fast transferred when the
main turbine generator was tripped.
There were other less significant operator distractions. Control rod drive number 39 was
noted to be sluggish moving a few inches out of synchronization with the other control*
rods in its group during withdrawal. Also, the temperature margin monitor low
temperature over-pressure pre-trip alarm had drifted slightly out of calibration low,
causing the annunciator to frequently alarm.
c.
Conclusions
The inspectors noted the operators were challenged by a number of emergent equipment
problems during the plant shutdown. This was indicative that the licensee continues to
struggle with significant plant material condition issues .
15
MS
Miscellaneous Maintenance Issues (92902 and 92700)
M8.1
(Closed) LER 95005-00: Inadvertent actuation of the safety injection system. On July 21,
1995, which the plant shut down for refueling, two inadvertent safety injection signals
were received while performing surveillance R0-12, "Containment High Pressure and
Spray System Test." In both instances, the left channel of safety injection was
inadvertently actuated. The cause of the first safety injection signal actuation was two
wires being taped together after removal from a terminal in an attempt to isolate the
safety injection system signal. The cause of the second safety injection signal aetuation
was a screwdriver inadvertently touching a terminal on the safety injection signal relay.
All equipment which had not been isolated prior to the event responded as required.
The licensee determined the root cause of the first inadvertent safety injection system
was inadequate communication between the system engineer procedure sponsor and
electrical maintenance personnel. This was the first time the revised procedure was used
and the first time electrical maintenance personnel performed the test. The licensee
determined that the second inadvertent actuation was due to insufficient work space.
The terminals were in a tight configuration where the wire was to be landed.
Procedure R0-12 was revised to specify that the wires removed from the terminal were to
be isolated and insulated from each other. caution statements were added to the
procedure regarding the tight work space, and suggested temporary insulation may be
necessary. Also, the training department collected test related problems that occurred in
the 1995 refueling outage for a case study lesson plan. This lesson plan was presented
to the engineering, operations and maintenance departments. This item is closed.
M8.2
(Closed) VIO 50-255/96008-03: Unauthorized operation of plant equipment. During
performance of emergency diesel generator preventive maintenance, an electrical
technician started the supply ventilation fan to the emergency diesel generator room
without authorization of operations personnel. The manipulation of the fan was contrary
to guidance in administrative procedure AP 4.02, "Control of Equipment." The electrical
technician's supervisor discussed the inappropriate action taken with the individual. The
various department managers, In standdown or continuous training meetings, discussed
with personnel the expectation that equipment control requirements in procedure AP 4.02
be followed. This item is closed.
M8.3
(Closed) LER 97004-00: Trip of high pressure safety injection pump while filling the
safety injection tank. On February 21, 1997, during boron concentration sampling of
safety injection tank T-82C, the tank was declared inoperable due to low level and
pressure. The licensee entered a one hour allowed outage time per TS 3.3.2.a to obtain
the sample. Concurrently, the high pressure safety injection pump P-66A tripped during
initial filling of tank T-82C. The trip of pump P-66A concurrent with tank T-82C being
inoperable was a condition prohibited by TSs, which required immediate entry into 3.0.3. *
Inspection of pump P-66A breaker by electricians found the Y-phase time over-current
relay target dropped. Root cause of the time over-current relay trip was a small metal
particle which lodged between the top surface of the induction disk and the relay's
magnet. This prevented the induction disk from rotating back to the reset position. The
particle was sent out for electron microscope examination. The laboratory analysis
determined the particle was a fragment to a terminal screw.
16
The time overcurrent relay was a sealed unit. This would lead to the conclusion that the
particle was introduced during manufacture. A new time over-current relay was installed
and tested satisfactorily. This item is closed.
M8.4
(Closed) LER 50-255/97005-00: Operation outside design basis due to unacceptable
Code repairs. Subsequent inspection after repairs to main steam isolation valves
CV-0501 and CV-0510, determined that unacceptable repairs had been made to the
valves' stuffing box pipe plugs which created steam leaks. These inadequate
- ASME Code violations were considered to have resulted in operation of the plant outside
the design basis. This event was in detail in Inspection Report No. 50-255/96017 and
50-255/97005. This issue resulted in violations (50-255/97005-01, 02, 03). The
corrective actions for this LER will be tracked under the violations; therefore this LER is
closed.
Ill. Engineering
E1
Conduct of Engineering
E1.1
Licensee Appendix R Review (37551)
a.
Inspection Scope
The inspectors reviewed the licensee's progress and recent findings that pertained to the
Appendix R enhancement effort. Two recent significant findings are detailed below. The
compensatory measures taken were reviewed for* adequacy. Discussions were held with
the system engineering supervisor and personnel supporting this effort. Applicable
documentation and procedures were reviewed.
b.
Observations and Findings
The licensee is presently working to complete the Appendix R enhancement effort and
complete incorporation of the 1996 refueling outage modifications into the Appendix R
data base. The applicable Appendix R analyses are being reviewed to be certain that all
conclusions of the supporting calculations have been incorporated into the base-Appendix
R engineering analyses. Affected off normal procedures are being revised. The licensee
identified the following Appendix R issues.
Design Issue One
On September 12, 1997, the licensee identified an error in a calculation assumption
during an Appendix R design bases review. Specifically, the error involved an improper
evaluation for effects that resulted in the spurious operation of valves due to the
introduction of a hot smart short in the electrical interface between the component cooling
water (CCW) and service water (SW) systems. A single spurious operation of a valve
between the interface of the CCW/SW systems could result in the loss of CCW inventory
to the lower pressure SW system. The time period in which the CCW inventory is lost
would not allow for manual action to prevent an unanalyzed condition.
17
The licensee made a 1-hour nonemergency 10 CFR 50.72 notification for being in a
condition outside of design bases. The part of the SW system involved was the seal
cooling water for the essential safety system pumps. The most limiting of the three
potential scenarios calculated that only 25 seconds would be available to close
engineering safeguards pump cooling service water retum valve, CV-0951. This valve is
normally closed. An open item of this calculation acknowledged the 25 second
requirement, but concluded since the essential safety system (ESS) pumps are not
running during normal operation and the CCW supply and retum valves (CV-0913 and
- CV-0950, respectively) to the ESS pumps' sealing cooling piping are normally closed,
then the spurious opening of any one CCW/SW interface valve could not result in the loss
of CCW inventory. Based on this information, another calculation did not consider
actions for the required time period. This reasoning is in error because CV-0913 and
CV-0950 are normally open and fail open on loss of air or loss of electric power. A single
spurious operation of CV-0951 would require a 25 second operator response to mitigate
the consequences, which is not possible. The licensee's compensatory measures are
detailed in Section 01.3 of this inspection report.
Design Issue Two
The second event was reported to NRC via a 10 CFR 50.72. It involved the Appendix R
analysis assuming all four primary coolant pumps being tripped if the fire causes an
evacuation of the control room. *The Off Normal Procedure for Alternate Shutdown did
not reflect the analysis and only directed the operators to trip two of the four primary.
coolant pumps.
The procedure ONP-25.2, "Alternate Safe Shutdown Procedure," does not specifically
address securing all the primary coolant pumps when the operators lose the ability to
monitor the pumps, such as during a control room evacuation or a damage to the
instruments. ONP 25.2 does not only cover fires where a control room evacuation is to
take place, but also provides guidance for fires where the control room is still manned.
The procedure assumes that monitoring of the primary coolant pumps is a condition of
their continued operation.
Several fire scenarios would result in a loss of component cooling water to the primary
coolant pump seals and bearing coolers. Upon leaving the control room, operators do not
have primary coolant pump monitoring capability or instrumentation to monitor the CCW
system. The licensee's design basis for the primary coolant pumps indicated they are
designed to operate without seal cooling for periods of up to ten minutes. Immediate
corrective action was to initiate a procedure revision to direct operators to trip all four
primary coolant pumps prior to control room evacuation.
This scenario assumes a fire is severe enough to cause a loss of CCW to the primary
coolant pumps via loss of CCW pumps or closure of the CCW valves to/from containment
but not severe enough to cause a loss of offsite power. Securing all primary coolant
pumps for a fire of lesser severity may not be prudent. For this type of scenario the
operators have additional guidance from ONP 6.2 "Loss of Component Cooling Water."
This procedure directs further securing of the primary coolant pumps for degraded CCW
18
cooling. This guidance is feH to be adequate in the short tenn until ONP 25.2 can be
clarified. Operator training includes the necessity of CCW cooling for primary coolant
pump for operation, lherefore the operators are encouraged to secure primary coolant
pumps when CCW is no longer capable of being monitored.
c.
Conclusions
The inspectors found the compensatory measures taken for the identified Appendix R
issues to be adequate. The Appendix R enhancement review was found to be
progressing slowly. However, the review appeared to be thorough.
EB
Miscellaneous Engineering Issues (92700 and 92903)
EB.1
{Closed) Licensee Event Report 95008-00: Bypassed containment high pressure trips on
reactor protection system. The licensee discovered that all four channels of containment
high pressure trip in the reactor protective system were inadvertently bypassed since a
modification in 1992. The event was the focus of Inspection Report No. 50-255/95010),
which resulted in violations 50-255/95010-01, 02 and 03. The corrective actions
documented in the LER will be tra~ed by the violations; therefore, this LER is closed.
EB.2
(Closed) LER 95012-00: Unqualified electrical connection in containment service water
outlet valve solenoid valve (SV-0824). The unqualified connection (wire nuts) was
located in a pull box in the component cooling water room outside eontainment. This
area is not affected by the loss of coolant accident pressure/temperature, but would be
exposed to radiation effects due to the radioactive "shine" through the containment wall.
The wire nut connections were replaced with environmentally qualified connections. A
walkdown of junction boxes to check for environmentally qualified connections done in
1992 missed this pull box because the electrical drawings used did not identify pull boxes.
Since the basis for the 1992 review were the electrical drawings the problem was not
discovered, and a 1995 review done after the identification of SV-0824 used the cable
and raceway schedule (E-33 series), which indicated the existence of a connection
regardless of the box type designation. The remaining pull boxes were inspected. All
remaining pull boxes identified on the E-33 series drawings were inspected, except four
which required the erection of scaffolding or were prohibitive due to high radiation
exposure. No other wire nut connections were found during this walkdown. Inspection of
the remaining four boxes is not considered necessary. This item is closed.
EB.3
(Closed) Violation 50-255/96-003-02: Failure to maintain design basis document (DBDs)
- current. The inspectors identified that the required biennial review and revision on 14 of
36 DBDs had not been perfonned with a 2-year period. During review of this issue,
engineering personnel identified several programmatic weaknesses and corrective
actions were initiated.
Engineering personnel developed a DBD change request log. Outstanding change
requests are now posted in front of controlled DBDs. All non-qualified DBD owners
- completed the necessary training. All DBDs were then subsequently reviewed to ensure
the DBDs were updated in a timely matter. The inspectors perfonned a random review of
DBDs. No over due change requests were identified.
19
E8.4
E8.5
E8.6.
The existing plant modification procedures specifically require the change initiator to
prepare a DBD change request. However, operating procedures do not cover this area.
This was addressed through a change to administrative procedure 3.07, "Safety
Evaluations," by the addition of the following question on the safety review form: "Should
this be included in a 080 update?" This item is closed.
(Closed) VIO 50-255/96008-04: Failure to follow surveillance procedure. On August 4,
1996, the licensee performed surveillance M0-7A-1, "Emergency Diesel Generators 1-1
and 1-2" for the emergency diesel generator 1-1. After the emergency diesel generator
was started, but prior to loading the generator, the system engineer noticed the arm of
the fuel rack was oscillating. The system engineer touched the fuel rack to stop the
motion. The surveillance continued and the diesel declared operable. This manipulation
of the fuel rack was contrary to the surveillance. The licensee re-performed the
surveillance on August 14, 1996, without any intervention with the functionality of the fuel
rack. The surveillance was completed satisfactorily. The system engineers supervisor
discussed the inappropriate action taken with the individual. The various department
managers, in standdown or continuous training meetings, discussed with personnel the
expectation that equipment control requirements in administrative procedure AP 4.02,
"Control of Equipment" be followed. This item is closed.
(Closed) LER 96010-00: On July 17, 1996, the safety injection tanks (SITs) were being
sampled for boron concentration. While filling SIT T-82C, high pressure safety injection
pump P-66A tripped. T-82C was inoperable due to low pressure during the sampling
evolution. With these two conditions, TS 3.0.3 was immediately entered. Two minutes
later, TS 3.0.3 was exited when T-82C nitrogen pressure was restored. A 24-hour
limiting condition for operation was then entered as required by TS 3.3.2.c.
Troubleshooting revealed that P-66A tripped due to the Y-phase time overcurrent (TOC)
relay on the pump breaker not resetting .after each successive start. The P-66A motor
was checked and cleaned. The TOC relay was* checked and calibrated. Then, P-66A
was test started three times. Proper resetting of the relay was verified after each start.
P-66A was declared operable, but degraded. On July 18, 1996, the LCO was exited.
The relay was replaced on July 1996.
Dust or grease buildup was determined to have caused an interference, preventing the
TOC relay from resetting. All similar relays were inspected for potentially degraded
conditions. No problems were found. This appeared to be a random failure. This item is
closed.
(Closed) Violation 50-255/96014-04: Inadequate design control of a temporary
modification to a polar crane solenoid. On November 6, 1996, temporary
modification 96-050 to the containment polar crane did not contain adequate installation
instructions for replacement of a single solenoid with two solenoids. The original solenoid
was hard mounted and was provided with sufficient ventilation t<>. prevent premature
failure. As a result of inadequate preparation and review, temporary modification 96-050
did not provide instructions for mounting the second of two replacement solenoids. The
second solenoid was installed using duct tape and tie-wraps in a manner which resulted
in overheating and failure of the solenoid coil, which resulted in a minor electrical fire .
20
The licensee subsequently replaced the two 250 volt coils with a new 460 volt coil. This
event was reviewed by the design engineering group and lessons learned were
developed. Discussions covered conditions leading to the event and the need to
consider all operating design characteristics. A review of all temporary modifications was
conducted to verify that acceptable standards were used for installation. Administrative
procedure 9.31, "Temporary Modification Control," section 7.2.3 was revised to add the
requirement to verify physical installation when critical to a design. This item is closed.
IV. Plant Support
R1
Radiological Protection
R1 .1
Spent Fuel Pool Resin Transfer
a.
Inspection Scope (71750 and 83750)
The inspectors observed the spent fuel pool resin transfer activity. Discussions were held
with the radwaste system engineer and supervisor on corrective actions instituted to
prevent resin inventory monitoring problems which occurred during the previous resin
sluice. The licensee received a violation for inadequate procedural controls for a
December 23, 1996 standard resin sluice from the purification and deborating ion
exchanger T-518 to spent resin storage tank T-100. Although there were several
contributing factors that led to the violation, the inspector's main concerns in observing
this resin transfer were the inadequate controls for monitoring tank resin levels and low
sensitivity to radwaste systems deficiencies by operations, system engineering and
radiation safety personnel.
b.
. Observations and Findings
The inspectors noted greater personnel resources were applied to the spent fuel pool
resin transfer than in similar previous activities. A senior reactor operator managed the
overall activities for the transfer of resin. A senior reactor operator was in constant direct
communication with the several auxiliary operators responsible for valve manipulations
and monitoring the radwaste system during the resin transfer. All the operators were
knowledgeable of their tasks. Health physics performed initial radiation surveys of the
tanks involved it the resin transfer, which were more detailed than previous surveys. The
flush of the resin transfer was also performed longer than in the past to ensure all resin
had been moved. Post radiation surveys of the tanks and in-line filters were also more
extensive than in past resin transfers.
The inspectors noted the procedure covering resin transfer had been revised to include
more thorough documentation of resin inventory. Precautions were added to the
procedure to heighten personnel awareness regarding the interconnecting system
flowpaths and the potential for abnormal radiological conditions such as plugging of
radwaste filters.
21
c.
Conclusions
The licensee's actions to improve the resin transfer process resulted in an error-free
evolution for the spent fuel pool job. The inspectors noted good attention to detail and
awareness of changing radiological conditions. Procedure revisions for tending resin
inventory and heightening personnel awareness of the radiological significance of
performing a resin transfer were successful.
V. Manaaement Meetings
X1
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management at
the conclusion of the inspection on October 17, 1997. No proprietary information was
identified.
22
. .
PARTIAL UST OF PERSONS CONTACTED
Licensee
R. A. Fenech, Senior Vice President,
Nuclear, Fossil, and Hydro Operations
T. J. Palmisano, Site Vice President - Palisades
G. B. Szczotka, Manager, Nuclear Performance Assessment Department
D. W. Rogers, General Manager, Plant Operations
D. P. Fadel, Director, Engineering
S. Y. Wawro, Director, Maintenance and Planning
R. J. Ger1ing, Manager, Design Engineering
P. D. Fitton, Manager, System Engineering
T. C. Bordine, Manager, Licensing
J. P. Pomeranski, Manager, Maintenance
D. G. Malone, Shift Operations Supervisor
M. P. Banks, Manager, Chemical & Radiation Services
K M. Haas, Manager, Training
M. E. Parker, Senior Resident Inspector, Palisades
P. F. Prescott, Resident Inspector, Palisades
23
u
INSPECTION PROCEDURES USED
IP 37551:
IP 61726:
IP 62707:
IP 71707:
Onsite Engineering
Surveillance Observations
Maintenance Observation
Plant Operations
IP 71750:
IP 83750:
IP 92700:
IP 92701:
Plant Support Activities
Occupational Radiation Exposure
Licensee Event Reports
Followup
IP 92901:
IP 92902:
IP 92903:
Followup - Operations
Followup - Maintenance
Followup - Engineering
ITEMS OPEN
"SO-~J97&!.1 01 -
-fcailyre to ~
50-255/97011-02
50-255/95005-00
50-255/95008-00
50-255/95012-00
50-255/96003-02
50-255/96008-03
50-255/96008-04
50-255/96010-00
50-255/96014-01a
50-255/96014-01b
50-255/96014-04
Failure to take adequate corrective actions for an Appendix R
concern
ITEMS CLOSED
LER
Inadvertent actuation of the safety injection
LER
Bypassed containment high pressure trips on reactor protection
system
LER
LER
Unqualified electrical connection in containment service water
Failure to maintain design basis document (OBOs) current
Unauthorized operation of plant equipment
Failure to follow surveillance procedure
Trip of high pressure safety injection pump
Failure to ensure control rod drive mechanisms locked prior to
inserting a reactor trip signal
Failure to position instrument AC bus Y-01 control handle to the
correct position resulting in loss of power to bus _ _
__
Inadequate design control of temporary modification for polar crane
solenoid
24
. ,
50-255/97004-00
50-255/97005-00
LER
Trip of high pressure safety injection pump while following the
safety injection tank
LER
Operation outside the design basis due to unacceptable Codes
repairs
25
A LARA
ccw
CFR
CROM
CV
ESS
gpm
IP
LCO
LER
MO
NRC
NCO
ONP
oos
QO
RV
sv
T
TS
WGS
UST OF ACRONYMS USED
As Low As Reasonably Achievable
Alternating Current
Administrative Procedure
American Society of Mechanical Engineers
Component Cooling Water
Code of Federal Regulations
Control Rod Drive
Control Rod Drive Mechanism
Control Valve
Design Basis Document
Division of Reactor Projects
Estimated Critical Position
Essential Safety System
General Operating Procedure
Gallons Per Minute
Inspection Procedure
Limiting Condition for Operation
Licensee Event Report
Monthly Operating (procedure)
Nuclear Regulatory Commission
Nuclear Control Operator
Non-Cited Violation
Off Normal Procedure
Out Of Service
Personnel Contamination Monitor
Primary Coolant System
Public Document Room
Quarterly Operations (procedure)
Refueling Operations (procedure)
Relief Valve
Safety Injection Tank
System Operating procedure
Tank
Time Over-Current
Technical Specification
Volume Control Tank
Violation
Waste Gas System
26