ML18067A689
| ML18067A689 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 09/05/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML18067A687 | List: |
| References | |
| 50-255-97-08, 50-255-97-8, NUDOCS 9709230082 | |
| Download: ML18067A689 (18) | |
See also: IR 05000255/1997008
Text
U.S. NUCLEAR REGULATORY COMMISSION
Docket No.:
License No.:
Report No.:
Licensee:
Facility:
Location:
Dates:
Inspector:
Approved by:
9709230082 970905
ADOCK 05000255
G
REGION Ill
50-255
50-255/97008(DRP)
Consumers Power Company
212 West Michigan Avenue
Jackson, Ml 49201
Palisades Nuclear Generating Plant
27780 Blue Star Memorial Highway
Covert, Ml 49043-9530
May 24 through July 7, 1997
P. Prescott, Resident Inspector
Bruce L. Burgess, Chief
Reactor Projects Branch 6
EXECUTIVE SUMMARY
Palisades Nuclear Generating Plant
NRC Inspection Report 50-255/97008
This inspection reviewed aspects of licensee operations, maintenance, engineering and
plant support. The report covers a 6-week period of resident inspection.
Operations
A plant procedure that allowed operations with steady state indicated reactor
thermal power greater than the licensed limit was identified as a violation. This
-procedure had previously been modified by the licensee to no longer allow the
steady state operation above the licensed limit. While no actual operation of the
unit of greater than the licensed limit was identified, the potential for such
operation had existed.
(Section 01 .2)
Ttie inspectors noted good operations performance during a CCW system spurious
relief valve lift. Operator identification and resolution of the event was prompt and
thorough. However, the inspectors identified a weakness in the initial operability
evaluation, which was subsequently addressed. The relief valve was subsequently
gagged closed until repairs could be initiated. (Section 01 .3)
Maintenance
The inspectors observed a weakness in communications in that neither system
engineering nor l&C personnel informed the operators of a grounding problem that
could occur during performance of the loop one T-ref maintenance activity, nor
were the alarms that could be received in the control room reviewed with the
operators. The inspectors noted these oversights were corrected in the loop two
phase of the maintenance activity. (Section M1 .2)
The inspector discussed an improper post maintenance test on valve CV-0733, and
indicated that this was another example of a negative trend observed in the quality
of post maintenance testing. The PMTs reviewed appeared to have been written to
verify the initial problem was repaired, not that the component continued to meet
its design function following maintenance. The licensee is currently reviewing the
PMT process. (Section M1 .3)
--- ---
-*-
2
~*
Engineering
The inspectors, in followup to a potentially generic issue, determined that the
licensee's administrative and design features that pertained to part length (P-L)
control rods provide sufficient controls such that a reactor power excursion due to
a stuck or mispositioned P-L control rod would be highly unlikely. Also, the
licensee's fuel vendor had reviewed and determined that a P-L control rod event
was bounded by a dropped or ejected control rod scenario in the current fuel cycle
analysis ~eport. (Section E1 .1)
Plant Support
The inspectors determined that the post maintenance critique did not fully address
.other available options to reduce dose during evaporation cleaning activities.
Critique meeting participants characterized the evaporator cleaning as a low dose
job (less than or equal to 10 mrem) when in fact the licensee had expended
approximately 350 mrem for a job that may not have been required. The inspectors
concluded that the evaporator cleaning job did not have the proper emphasis placed .
on ALARA planning. (Section R 1. 1)
3
REPORT DETAILS
Summary of Plant Status
The plant operated at essentially 99.6 percent power for the entire inspection report
period. July 7, 1997, marked the 138th day of continuous power operation.
I. Operations
01
Conduct of Operations .
01.1
General Comments (71707)
01.2
a.
b.
- Using Inspection Procedure 71707, the inspectors conducted frequent reviews of
ongoing plant operations. The inspectors considered the conduct of operations to
be good. Specific events and noteworthy observations are detailed below.
Followup on Exceeding Licensed Thermal Power Limits
Inspection Scope (71707)
During this inspection period, the NRC completed its review of enforcement action
(EA)96-092 concerning a February 7, 1996 event at Palisades involving the
potential to exceed rated reactor thermal power limits as indicated by available
control room power monitors. Inspection report 50-255/96002(DRP) provided the *
specific facts and. preliminary analysis of this event. Below is a discussion of the
NRC's review and conclusions concerning the licensee's operation at near full
power.
Observations and Findings
On February 7, 1996, reactor thermal power was indicated to have exceeded the
power stated in the facility's license. This inadvertently occurred during a
delithiation evolution to control primary coolant system chemistry parameters. The
operations shift was aware that, by procedure GOP-12, Revision 12, reactor power
was allowed to reach 100.99 percent. Reactor thermal power is measured by
nuclear instrumentation that is calibrated periodically using a heat balance
calculation. A heat balance calculation provides the best indication of actual
reactor thermal power. Accident analyses presented in the FSAR must meet the
requirements of 10 CFR 50 Appendix K "ECCS Evaluation Models." These analyses
are performed assuming a reactor thermal power of 102 percent in order to allow
for instrument uncertainties. By exceeding licensed thermal power limits, reactor
- - powe{ during an accident scenario could potentially be outside design bases
.
because the margin of safety derived from assuming a 2 percent instrument error
would be reduced by the higher initial power level at the time of accident initiation .
4
-*
c.
However, the inspectors determined the safety significance of this event was
minimal. Review of subsequent tests and analyses showed that the licensee did
not exceed 1 00 percent power during the nine hour delithiation process. A
calorimetric uncertainty analysis was completed that utilized instrumentation and
indication uncertainties and an ultrasonic flow measurement (UFM) of the feedwater
flow rate was performed. The UFM provided a more accurate indication of actual
feedwater flow, independent of the installed feed water venturies. Due to
feedwater flow rates being the single largest contributor in a calorimetric
calculation; small errors in feedwater flow rates could result in larger differences in
indicated reactor power. Results of the UFM testing revealed that actual power
was 2.2 percent less than the indicated power based on use of the feedwater flow
venturies. The difference was due to a conservative initial venturi calibration and
venturi fouling. Using the UFM results, maximum power level achieved during the
.delithiation process was determined to be 98.2 percent.
NRC has issued guidance that licensees may not operate above the steady state
indicated reactor thermal power limits stated in the license, except in unanticipated
transient conditions. If steady state indicated reactor thermal power exceeds the
licensed limit, the guidance directed licensees to initiate prompt corrective action
within 15 minutes to restore reactor power to less than or equal to the license
power limit.
The inspectors wrote a task interface agreement (TIA) issued to the Office of
Nuclear Reactor Regulation (NRR) to evaluate the adequacy of existing guidance.
The basis for the TIA was that present technology allows c~lculating, almost
instantaneously, reactor thermal power. The current standing guidance to the
industry was developed when calculating reactor thermal power was a one hour or
longer process. Thus under the old technology, the delithiation process and
resultant indicated power level of over 1 00 percent would not have been
immediately detected. Using current technology, almost anytime any evolution
re1ises power above 100 percent, the power excursion would be detected and raise
a question regarding whether or not a licensee should perform a calorimetric
knowing an overpower indication exists.
The response to the TIA stated that the deliberate raising of power above the
licensed limit was inappropriate. Procedure GOP-12, Revision 12, allowed the brief
operation in excess of licensed reactor thermal power. This procedure was
inappropriate to the circumstances and is considered a violation of 10 CFR 50
Appendix B, Criterion V "Instructions, Procedures, and Drawings," (50-255/97008-
01 (DRP)).
In response to NRC concerns, licensee management modified the procedure such
. _ that-it no longer allowed steady state power operation above the licensed limit.
Conclusions
A plant procedure that allowed operations with steady state indicated reactor
thermal power greater than the licensed limit was identified as a violation. In
5
response to concerns from the NRC, licensee management modified the procedure.
While actual operation of the unit greater than the licensed limit was not identified,
the potential for such operation had existed.
01.3
Component Cooling Water (CCW) Relief Valve (RV> Lift During Surveillance
a.
Inspection Scope (71707. 61726 and 37551)
The inspectors observed operations personnel conduct a prejob brief and perform a
right channel surveillance using procedure 00-1, "Safety Injection System."
b.
Observations and Findings
.The purpose of surveillance procedure 00-1 was to demonstrate operability of the
right channel of the safety injection system (SIS) initiation circuitry (SIS actuation
relays and design basis accident (OBA) sequencer) by using the internal testing
capability of the system. One system tested is the component cooling water
(CCW) system. The SIS initiation circuitry signals one of the other two CCW
pumps to start (one normally is already in service). During performance of 00-1 on
June 9, 1997, CCW pump P-528 automatically started as required. This resulted in
an expected increase in CCW system pressure. However, relief valve RV-2108,
which provides thermal over pressure protection for the shield cooling heat
exchanger, subsequently lifted .
The valve did not reseat normally which resulted in an approximately two gpm leak.
No alarms are automatically actuated when relief valve RV-2108 lifts, thus the
operating crew was not immediately aware of the partially open valve. An extra
nuclear shift operator (NSO) was assigned to assist in the control room while the
two normal onshift NSOs performed 00-1. During a routine panel walkdown, the
extra NSO noticed a decrease of approximately 10 percent in the CCW surge tank
level. The extra NSO also noted to the control room supervisor that containment
sump level was trending up. The operators checked the volume control tank level
to verify there was no decrease in level and to ensure that a primary coolant
system leak had not occurred. The operators then concentrated on finding a CCW
leak.
The operators:
Calculated CCW surge tank level loss to determine the rate of decrease;
stopped testing of 00-1;
. *--restored-the plant to normal configuration following the suspension of
surveillance test 00-1 ; and
entered the off-normal procedure for the CCW system due to the apparent
leak.
6
~.
The off-normal procedure was reviewed by operations personnel and the location of
all relief valves on a CCW system drawing were identified. Also, personnel on a
standby list of maintenance and system engineering personnel were notified. A
CCW corrective action team entered containment and identified that RV-2108 for
the shield cooling system had lifted and stayed opened. The valve was
mechanically agitated and it subsequently reseated. Licensee personnel generated a
condition report and an initial operability evaluation was performed. Operators
noted that the available indicators for the relief valve indicated that the valve lifted
early since when the relief valve lifted, CCW pressure was approximately 135 psig
and the setpoint of the relief valve was 150 psig.
The inspectors noted good operator response to the stuck open relief valve and
small CCW leak inside of containment.
The inspectors identified one weakness with the initial operability evaluation.
Initially, the evaluation addressed only the as found leak rate of 2 gpm and failed to
address the potential leak rate of a full open relief valve. If RV-2108 had lifted to
its full capacity of 24 gpm, the inspectors were concerned that the CCW surge tank
makeup capability would be insufficient. System engineering calculated that the
makeup capability of the CCW system was 150 gpm, which would be sufficient to
maintain the CCW system operable should RV-2108 spuriously lift again.
Subsequently, the valve was gaggeo closed to prevent recurrence. Two other relief
valves associated with the CCW system were verified to provide adequate
protection for the shield cooling heat exchanger from over pressure until RV-21 08
can be replaced.
c.
Conclusions
The inspectors noted good operator performance during identification and response
to the spurious lift of a CCW relief valve. Operator identification and response to
restore CCW system integrity was prompt and thorough. However, the inspectors
identified a weakness in the initial operability evaluation, which .was subsequently
addressed. The relief valve was subsequently gagged closed until repairs can be
initiated.
II. Maintenance
M 1
Conduct of Maintenance
M 1 . 1 General Comments
a.
Inspection Scope !62707 and 61726)
The inspectors observed all or portions of the following work activities:
Work Order No:
24711110
Dirty waste "B" evaporator; open/inspect and hydrolaze
7
24711266
CV-3223, SOC HXH E-60A inlet valve; open/inspect
PCV and replace internals
24711268
. CV-321 2, SOC HXH E-608 inlet valve; open/inspect
PCV and replace internals
24711416
CV-3055, inlet valve to SOC HXH; open/inspect PCV
and replace internals
24711267
CV-3224, SOC HXH E-60A outlet valve; open/inspect
PCV and replace internals
24514371
Install new program for PCS Loop one revised T-ref
curve in transmitter TYT-0100 per SC-95-099
24612597
Install new program for LIC-01 OA pressurizer level
controller for revised T-ref curve on loop one
24612508
CV-0511 turbine bypass valve; replace tubing and
fittings downstream of CA-0390
24513316
Diagnostic testing of CV-0511
24514370
Install new program for PCS loop 2 revised T-ref curve
in transmitter TYT-0200 per SC-95-099
24612596
Install new program for LIC-0101 B pressurizer level
.~
~
controller for revised T-ref curve on loop two
~
24612911
Charging pump P-55A; install new pump body and head
24712354
Hydrolaze drain line to equipment drain tank T-80
Surveillance Activities
SOP-2
. Surveillance for Auxiliary Feedwater valves CV-
0727 and CV-0749 following PPAC FWS034
SOP-8 ATT 2
Testing of Main Turbine Valves/Protective Trips
00-1
Safety Injection System (Right Channel With
Standby Power)
00-1
Safety Injection System (Right Channel Without
Standby Power)
00-19
lnservice Test Procedure - HPSI Pump and ESS
Check Valve Operability Test
8
-*
b.
Observations and Findings
The inspectors concluded that the work performed during maintenance and
surveillance activities was professional and thorough. All work observed was
performed with the work package present and in active use. Work packages were
comprehensive for the task and post maintenance testing requirements were
adequate. The inspectors frequently observed supervisors and system engineers
monitoring work practices. When applicable, work was completed by adhering to
the appropriate radiation control practices.
c.
Conclusions
In general, the inspectors observed good procedure adherence, maintenance and
.radiation worker practices. Specific observations are detailed below.
M1 .2 Poor Communications During T-ref Controller Maintenance
a.
Inspection Scope (61726 and 71707)
The Inspectors observed portions of scheduled maintenance on transmitters TYT-
0100 and TYT-0200. The temperature reference (T-ref) curve had changed and the
licensee intended to change the electronic program constants to reflect the revised
curve.* In addition to observing the transmitter work, the inspectors also reviewed
the associated work package and observed the post maintenance test. Also
observed were maintenance activities for the pressurizer level controller LIC-0101 A
and LIC-0101 B, which provided a revised pressurizer level setpoint curve. The
pressurizer level setpoint curve was revised to reflect a revised Tave for 100 percent
power.
b.
Observation and Findings
As noted in section 01.2 of this report, the licensee had identified conservative
errors in the measured flow rates of the main feedwater system. Following the
identification of these errors, l&C personnel adjusted feedwater flow
instrumentation and other power measuring instruments. As the unit power was
adjusted, Tave and T-ref were also adjusted.
The first portion of the maintenance activity involved removal of the loop one
transmitter TYT-0100 to have its program upgraded and then reinstalled after
testing. TYT-0100 was unplugged from the control room panel and a digital
programmer was connected. When the programmer was turned on and TYT-0100
was plugged back in an AC ground fault alarm occurred on preferred AC power bus
.. 'l~10, which powers TYT-0100. The TYT-0100 showed no sign of having AC
power applied. Also, digital T.v. indicator Tl-0111 and temperature recorder TR-021
both showed a 15° F increase. At this point, the control room operators suspended
the job and entered the proper annunciator response procedure. The inspectors
observed good command and control of control room operations.
9
The ground fault was evaluated and the required procedural actions completed.
The operators then allowed removal of transmitter TYT-0100. The ground was no
longer observed on the Y-10 bus. The original TYT-01 00 transmitter unit was
replaced with a new unit. The inspectors learned from discussions with the system
engineer that similar events had occurred with the same model transmitters in five
previous instances. The inspectors had attended the prejob brief and this potential
problem was not discussed. The inspectors also noted that the operations
personnel were not present for the prejob brief. Prior to commencing work, neither
the system engineer nor instrumentation and control (l&Cl technieians briefed
operations of this potential problem. During this evolution, the inspectors discussed
with plant management concerns that operators are briefed on expected alarms
prior to commencement of work .
. Prior to work on the second Tave loop, the inspectors discussed with operations that
LIC-0101 B was a suspect unit and that the scope of the job was to only reprogram
the unit. The operators performed a prejob brief for the loop two work activity with
the operations shift, l&C technicians, their supervisor, and the system engineer.
The inspectors noted the brief was thorough. During the brief, the system engineer
identified to operations that LIC-0101 B was a suspect unit. The original scope of
the work package was to simply reprogram the unit and not replace the unit or the
power supply. The operators suggested that it would be prudent to take care of
the potential power supply problem now rather than simply reinstall the unit. The
system engineer agreed and the power supply was replaced after proper work order
revisions were completed.
c.
Conclusions
The inspectors observed that neither system engineering nor l&C personnel
informed the operators of a potential problem that could occur during performance
of the T-ref maintenance activity, nor were potential alarms reviewed with
operations. The inspectors noted these oversights were corrected in the second
phase of the maintenance activity.
M-1.3 Adequacy of Post Maintenance Test (PMTI Requirements
a.
Inspection Scope 62707
The inspectors observed portions of maintenance performed for turbine bypass*
valve CV-0511 and portions of the testing conducted on condensate fast makeup
valve, CV-0733.
The PMT history for the CV-0733 valve was also reviewed.
b.
Observations and Findings
The intent of the work order for CV-0511 was to replace a mix of copper and
stainless steel instrument air lines and fittings with new stainless steel. Part of the
work order required removal of certain solenoid valve (SVs). The SVs were to be
de-terminated and the wire-nutted connections replaced with lugged connections.
10
In the inspectors' review of the PMT operability requirements for CV-0511, CV-
0511 was to remain isolated from the main steam system during valve timing tests.
Also, the PMT required the verification of no air leakage on the replaced instrument
air lines. The inspectors were concerned this would be an inadequate PMT of CV-
0511 , in that fill the SVs would not be verified as functional. The SVs for the
bypass valve quick opening and loss of condenser vacuum functions would not be
tested. The turbine bypass valve is important to plant safety. The valve passes up
to 4.5 percent steam flow with the reactor at full power. The FSAR states that the
turbine bypass valve is one of the systems utilized for taking the plant to hot
shutdown. The valve is also discussed in the Technical Specification Basis Section 2.2. The TS basis states that additional assurance is provided by the
bypass valve in preventing the nuclear steam supply system pressure from
exceeding safety limits.
The inspectors observed the PMT for the condensate fast makeup valve. The 12
inch valve supplies condensate makeup from storage tank T-2. The valve opens on
a low-low hotwell level signal. A pressure control valve (PCV) was repaired and
operators attempted to stroke CV-0733. The valve failed to stroke. The PCV was
adjusted to increase air pressure and the valve was mechanically agitated.
However, the valve still failed to stroke. The valve maintenance supervisor stopped
further testing until the situation could be reviewed. At the end of the inspection
period, the valve had not been stroked. Licensee maintenance personnel are
evaluating options to stroke the valve.
The licensee, in reviewing post maintenance and interviews with personnel, found
that the actuator for CV-0733 had been overhauled in February 1997. After
completion of the overhaul, it was found that CV-0733 had not been stroked even
though the PMT cover sheet recommended it. The licensee wrote a condition
report which requires a root cause analysis (referred to as a level two condition
report).
In the previous inspection report (IR) (50-255/97006(DRP)), the inspectors
identified a concern with PMT of the P-55A charging pump. Also, detailed in the
same IR were problems of PMT with P-88 Auxiliary feedwater pump. The
inspectors discussed with the licensee the continued weaknesses noted in the area
of post maintenance testing.
c.
Conclusions
The licensee reviewed the test requirements and decided to stroke the valve and
declare CV-0511 inoperable but available pending testing of the SVs and associated
control circuitry for the loss of vacuum and turbine trip features. The licensee
---discussed -the adequacy of the testing requirements with the work order planning
group. The licensee is currently reviewing the best method to stroke CV-0733.
The inspectors discussed the negative trend in post maintenance testing with the
licensee. The PMTs appeared to have been written to verify the initial problem was
repaired, not that the component continued to meet its design function following
11
maintenance. In response to NRC questions, regarding the PMT program, the
licensee is currently reviewing the PMT process.
Ill. Engineering
E1
Conduct of Engineering
E1 .1
Review of Part-Length Control Rod Transient Analyses
a.
Inspection Scope (37551 >
The inspectors reviewed the applicability to Palisades of a generic NRC concern
with part-length (P-L) control rods. Specifically, the concern involved the fuel
.vendor's elimination of two transient analysis events from the fuel cycle-by-cycle
analysis normally performed for the Combustion Engineering (CE) plants that have
the core protection calculator (CPC) digital protective systems.
To assess the applicability of this concern to the licensee and to verify any
necessary corrective actions the inspectors held discussions with reactor
engineering and operations department personnel. In addition, the inspectors also
reviewed licensee plant procedures, Technical Specifications (TS), the Final Safety
Analysis Report and operator training guides.
b.
Observations and Findings
The issue involved not addressing certain control rod misoperation events. The two
accident analysis of interest involved P-L control rod deviations while in the control
deadband during startup and the slip of a P-L control rod from 50 percent inserted
to 90 percent inserted. NRC review of this issue concluded that a single P-L control
rod deviation within the deadband and the P-L control rod slip are an anticipated
operational occurrence (AOO). An'AOO is an event in which plant conditions may
be present for this event once in the life of the plant. Therefore, it must be
evaluated each cycle under all conditions allowed by TS.
The inspectors found several distinctions in plant configuration and administrative
controls that make the two events in question highly unlikely at Palisades.
( 1)
Palisades has 20 shutdown, 21 regulating and 4 P-L control rods. The P-L
control rod drive mechanisms (CRDMs), unlike the other CRDMs at
Palisades, has a short drive shaft in place of a clutch. The CROM motor and
brake cannot be uncoupled from the control rod without disassembly. As a
result, P-L ,control rods cannot drop into the core on a reactor trip, unlike
--- other CE plants with the digital protective systems ..
(2)
At Palisades, the rods in question are not used for flux shaping during power
operations. Licensee TS require P-L control rods to be completely withdrawn
from the core (except for control rod exercises and physics tests).
Administratively, P-L control rods are not exercised. Also, a P-L control rod
12
is considered inoperable if it is not fully withdrawn from the core and cannot
be moved by .its operator. By TS, if more than one control rod or P-L control
rod becomes misaligned or inoperable, the reactor shall be placed in the hot
shutdown condition within 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The licensee placed these restrictions
on P-L control rods since it had been previously demonstrated on other CE
plants that design power distribution envelopes could, under some
circumstances, be violated by using P-L control rods.
The inspectors discussed with reactor engineering potential operator error
scenarios. An operator could move P-L rods, but procedures do not allow it, except
during startup prior to criticality. The operator would have to commit two errors to
move a P-L rod. The operator would have to move the group selector switch for P-
L rods, which gives an alarm, then move the joystick that would move the rod.
This would also give an alarm. These actions would also be contrary to operator
training.
The inspectors reviewed licensee surveillances pertaining to control rod movements.
The procedures did not allow movement of P-L control rods.
c.
Conclusions
The inspectors determined that the licensee's administrative and design features
that pertained to part length (P-L) control rods provided sufficient control such that
a reactor power excursion due to a stuck or mispositioned P-L control rod would be
highly unlikely. Also, the licensee's fuel vendor had reviewed and determined that
a P-L control rod event was bounded by a dropped or ejected control rod scenario in
the current fuel cycle analysis report.
IV. Plant Support
R 1
Radiological Protection
R 1.1
Maintenance Activities and Daily Radiological Work Practices
a.
Inspection Scope (71750 and 83750)
The inspectors observed radiological worker activities during various maintenance
activities detailed in this inspection report, and also monitored radiological practices
during daily plant tours.
b.
Observations and Findings
- The inspectors' observation of jobs in progress during the maintenance activities.
detailed above revealed that radiation protection technicians were visible at the job
sites. The technicians took appropriate actions and surveys in accordance with
good ALARA practices.
13
c.
Conclusions
The inspectors concluded that radiological practices observed during the
maintenance activities and plant daily walkdowns were adequate. The inspectors
had no concerns. Specific observations are detailed below.
R1 .2 ALARA Planning of "B" Radwaste Evaporator Maintenance
a.
Inspection Scope (71750 and 83750)
The inspectors observed maintenance activities for the opening, inspecting and
cleaning of the B rad waste evaporator.
b.
-Observations and Findings
The inspectors attended meetings and held discussion with ALARA planning,
system engineering and other personnel involved with the B evaporator
maintenance task. The inspectors' main focus was on ALARA practices for
cleaning evaporator internals. A post maintenance critique meeting was also
observed.
The inspectors noted that this job had potential for significant dose accumulation
and that good ALARA planning and interdepartmental communication would be.
required to achieve a low total dose. Maintenance on the evaporator was being
performed because of the overall poor material condition of the-evaporator system.
Auxiliary components were also scheduled for maintenance, besides cleaning
evaporator internals. In addition, operations viewed system performance and
reliability as poor.
,'
Engineering had determined that cleaning evaporator internals would improve
system performance. The method the licensee chose to clean the evaporator was
hydrolazing. Based on past experience, the inspectors questioned why a citric acid
flush of the evaporator was not considered. The licensee responded that although
a citric acid flush was considered, it had not been fully evaluated. The assumption
was environmental engineering would disapprove a citric acid flush because of the
amount of mixed radwaste generated.
After accessing the evaporator internals the licensee began to hydrolaze. Due to
the construction of the evaporator, the licensee found that most of the internals
were not accessible for hydrolazing. At the outset of th_e work, the inspectors
asked system engineering for work specific drawings. System engineering
responded that no drawings were available and the specific vendor was thought to
no long-er exist. -The inspectors, through discussions with the licensee, --*- -
-- ----
found cognizant individuals within the licensee's organization who were
knowledgeable of the vendor and confirmed the inspector's supposition that the
vendor still existed.
14
A contractor engineer had discussed how many evaporator steam tubes could be
plugged. This information was not relayed rad protection to the engineers or
ALARA coordinators planning the job. Although detailed drawings of the
evaporator were not available, vendor personnel responsible for the original
installation relayed important and previously misunderstood operational information
that would have made a detailed inspection of the evaporator internals unnecessary
prior to the planned maintenance.
The inspectors noted in the post maintenance critique a failure to fully investigate
the feasibility of a citric acid cleaning, especially after the problems encountered
with hydrolazing. The total dose expended for the evaporator internals inspection
and cleaning was approximately 350 mrem. The maintenance window to allow for
the reconditioning of the entire B evaporator system was large - at least three
.weeks. Even after the licensee knew the vendor still existed, the job continued as
originally planned without considering the information supplied by the vendor.
c.
Conclusions
The inspectors determined that the post maintenance critique did not fully address
other options available to reduce dose during evaporator cleaning activities.
Critique meeting participants characterized the evaporator cleaning as a low dose
job (less than or equal to 10 mrem) when in fact the licensee had expended
approximately 350 mrem for a job that may not have been required. The inspectors
- concluded that the evaporator cleaning job did not have proper emphasis placed on
A LARA planning.
V. Management Meetings
X1
Exit Meeting Summary
The inspectors presented the inspection results to members of licensee
management at the conclusion of the inspection on July 7, 1997. No proprietary
information was identified.
15
PARTIAL LIST OF PERSONS CONT ACTED
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 of Engineering
S. Y. Wawro, Director, Maintenance and Planning
J. L. Hanson, Director, Strategic Business Issues
R. J. Gerling, Design Engineering Manager
A. L. Williams, Acting 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
IP 37551:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 83750:
INSPECTION PROCEDURES USED
Onsite .Engineering
Surveillance Observations
Maintenance Observation
Plant Operations
Plant Support Activities
Occupational Radiation Exposure
ITEM OPENED
50-255/97008-01
Exceeding licensed thermal power limits
None
AVE
ccw
CFR
CROM
CV
GL
GPM
HXH
l&C
MREM
Mwt
NRC
NSO
ITEMS CLOSED
LIST OF ACRONYMS USED
As Low As Reasonably Achievable
Anticipated Operational Occurrence
Average
Component Cooling Water
Combustion Engineering
Code of Federal Regulations
Core Protection Calculator
Control Rod Drive Mechanism
- Control Valve
Design Basis Accident
Division of Reactor Projects
Enforcement Action
. Emergency Core Cooling System
Final Safety Analysis Report
Generic Letter
Gallons per minute
General Operating Procedure
High Pressure Safety Injection
Heat Exchanger
Instrumentation & Control
Level Instrument Controller
Milli~Rem
Megawatts Thermal
Nuclear Regulatory Commission
Nuclear Regulatory Research
Nuclear Shift Operator
Primary Coolant System
Pressure Control Valve
Public Document Room
..
P-L
PPAC
RV
sv
Tl
T-ref
TR.
TS
TYT
UFM
~ --- --- -----
Part-Length
Periodic & Predetermined Activity Control
Relief Valve
Safety Injection System
Temperature Indicator
Task Interface Agreement
Temperature - Reference
Temperature Recorder
Technical Specification
Temperature Transmitter
_Ultrasonic Flow Measurement
Violation