ML20151K541
| ML20151K541 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/12/1988 |
| From: | Rebecca Barr, Mendonca M, Suh G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20151K510 | List: |
| References | |
| 50-344-88-24, NUDOCS 8808030208 | |
| Download: ML20151K541 (12) | |
See also: IR 05000344/1988024
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U.S.-NUCLEAR REGULATORY COMMISSION
REGION V
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. Report No.
50-344/88-24
Docket No.
50-344
License No.
Licensee:
Portland General Electric Company
121 S.W. Salmon Street
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Portland, OR 97204
Facility Name: Trojan
Inspection at:
Rainier, Oregon
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Inspection condu ed:
8-J[ 18, 1988
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Inspectors:
M. C. Barr(/ "
Date signed
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Approved By:
K./ M. Mend
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Date Signed
Reactor Pr
ts Section 1
Summary:
Inspection on May 8 - June 18,-1988 (Report 50-344/88-24)
Areas Inspected:
Routine inspection of operational safety verification,
maintenance, surveillance (containment incal leak rate testing), event follow
up, and open item follow up.
Inspectic.1 procedures 30703, 41400, 61720,
62703, 71707, 71709, 71881, 92700, 92701 and 93702 were used as guidance
during the conduct of the inspection.
Results:
In the areas inspected, a violation was identified for an apparent failure to
follow procedural requirements for work hour limitations (paragraph 6).
No
general conclusions regarding the strengths or weaknesses of the program areas
inspected were identified during this inspection period.
8808030208 880713
ADOCK 05000344
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DETAILS
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1.
Persons Contacted
- D.W. Cockfield, Vice President, Nuclear
- C.A. Olmstead, Plant General Manager
- 'R.P. Schmitt, Manager, Operations and Maintenance
- D.W. Swan, Manager, Technical Services
- J.K. Aldersebaes, Manager, Plant Modifications
- J.D. Reid, Manger, Plant Services
- J.W. Lentsch, Manger, Personnel Protection
R.L. Russell, Operations Supervisor
R.H. Budzeck, Assistant Operations Supervisor
D.L. Bennett, Maintenance Supervisor
R.A. Reinart, Instrument and Control Supervisor
T.O. Meek, Radiation Protection Supervisor
R.W. Ritschard, Security Supervisor
C.H. Brown, Operations Branch Manager, Quality Assurance
The inspectors also interviewed and talked with other licensee
employees during the course of the inspection.
These included shift
supervisors, reactor and auxiliary operators, maintenance personnel,
plant technicians and engineers, and quality assurance personnel.
- Denotes those attending the exit interview.
2.
Plant Status
The plant was in Mode 6 with fuel assemblies being reloaded into the
reactor core at the beginning of the inspection period.
Near the end of
the previous inspection period on May 5, corporate engineering informed
the plant that the pressurizer surge line was found to be in contact with
one of its pipe whip restraints under cold conditions.
The licensee
initiated a detailed action plan to respond to pipe whip restraint and
pressurizer surge line operability concerns.
On May 13, core reload was
completed.
Following setting of the head on the reactor vessel, reactor
coolant system level was reduced to loop centerline on May 16 in
preparations for eddy current testing of the "A" and "D"
steam
generators.
Following opening of the steam generator primary manways and
installation of nozzle dams, the RCS level was increased to above the top
of the loops.
On May 23, the plant entered Mode 5 as the reactor vessel
head bolts wero tensioned.
On May 27, ultrasonic examination of
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pressurizer surge line welds resulted in relevant indications on the two
welds associated with the 90 degree elbow located below the pressurizer.
The licensee subsequently decided to replace the elbow and adjoining
pieces of piping.
On June 8, reduction of RCS level to slightly above
loop centerline was initiated to support the pressurizer surge line elbow
replacement work.
At the end of the inspection period, the plant was in
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Mode 5.
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3.
Operational Safety Verification (71707, 71709, 41400 and 71881)
During this inspection period, the inspectors observed and examined
activities to verify the operational safety of the licensee's facility.
The observations and examinations of those activities were conducted on a
daily, weekly or biweekly basis.
Daily the inspectors observed control room activities to verify the
licensee's adherence to limiting conditions for operation as prescribed
in the facility Technical Specifications.
Logs, instrumentation,
recorder traces, and other operational records were examined to obtain
information on plant conditions, trends, and compliance with regulations.
On occasions when a shift turnover was in progress, the turnover of
information on plant status was observed to determine that pertinent
information was relayed to the oncoming shift personnel.
Each week the inspectors toured the accessible areas of the facility to
observe the following items:
(a) General plant and equipment conditions.
(b) Maintenance requests and repairs.
(c) cire hazards and fire fighting equipment.
(d) Ignition sources and flammable material control.
(e) Conduct of activities in accordance with the licensee's
administrative controls and approved procedures.
(f)
Interiors of electrical end control panels.
(g)
Implementation of the licensee's physical security plan.
(h) Radiation protection controls.
(1) Plant housekeeping and cleanliness.
(j) Radioactive waste systems.
(k) Proper storage of compressed gas bottles.
Weekly, the inspectors examined the licensee's equipment clearance
control with respect to removal of equipment from service to determine
that the licensee complied with technical specification limiting
conditions for operation.
Active clearances were spot-checked to ensure
that their issuance was consistent with plant status and maintenance
evolutions.
Logs of jumpers, bypasses, caution and test tags were
examined by the inspectors.
Each week the inspectors conversed with operators in the control room,
and with other plant personnel.
The discussions centered on pertinent
topics relating to general plant conditions, procedures, security,
training and other topics related to in progress work activities.
The inspectors examined the licensee's nonconformance reports (NCRs) to
confirm that deficiencies were identified and tracked by the system.
Identified nonconformances were being tracked and followed to the
completion of corrective action.
Routine inspections of the licensee's physical security program were
performed in the areas of access control, organization and staffing, and
control measures used at the entrance to the protected area, verified the
integrity of portions of the protected area barrier and vital area
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barriers, and observed in several instances the implementation of
compensatory measures upon breach of vital area barriers.
Portions of
the isolation zone were verified to be free of obstructions.
Functioning
of central and secondary alarm stations (including the use of CCTV
monitors) was observed.
On a sampling basis, the inspectors verified
that the required minimum number of armed guards and individuals
authorized to direct security activities were on site.
The inspectors conducted routine inspections of selected activities of
the licensee's radiological protection program.
A sampling of radiation
work permits (RWP) was reviewed for completeness and adequacy of
information.
During the course of inspection activities and periodic
tours of plant areas, the inspectors verified proper use of personnel
monitoring equipment, observed individuals leaving the radiation
controlled area and signing out on appropriate RWP's, and observed the
posting of radiation areas and contaminated areas.
Posted radiation
levels at locations within the fuel and auxiliary buildings were verified
by the inspectors using both NRC and licensee portable survey meters,
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The involvement of health physics supervisors and engineers and their
awareness of significant plant activities was assessed through
conversations and review of RWP sign-in records.
The inspectors verified the operability of selected engineered safety
features.
This was done by direct visual verification of the correct
position of valves, availability of power, cooling water supply, system
integrity and general condition of equipment, as applicable.
No violations or deviations were identified.
4.
Maintenance (62703)
Cable Insulation Degradation Near Reactor Coolant Pumps
As a result of electrical preventive maintenance performed on various
valves associated with the reactor coolant pumps, the licensee identified
apparent degradation of their wiring insulation.
The most severe
degradation appeared to be localized to valve components where the wiring
was not protected by a cable outer jacket.
In these cases, portions of
some wiring showed insulation embrittlement.
A program to determine the
extent and cause of the apparent degradation was implemented.
This
program involved examination of cabling in open cable trays and of
cabling accessible upon removal of condulet covers, pull box covers,
terminal box covers, and in some instances disassembly of conduit
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fittings.
Much of the cable degradation wa; associated with non-safety
related cabling and involved various degrees of embrittlement of the
cable outer jacket.
With few exceptions, the enclosed conductors and
associated insulation showed no damage.
This program determined that the
problem was associated with high ambient temperatures caused by air flow
from reactor coolant pump motors.
The licensee developed an action plan
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which included repair and replacement of safety related wiring which
showed degradation, preparation of a safety evaluation which provided the
bases for continued operation with non-safety related cabling which
showed various degrees of degradation, and preparations for potential
replacement of degraded non-safety related cabling in the future.
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addition, the licensee contacted the NSS$ vendor and initiated efforts to
assess the generic implications of the observed conditions.
The inspectors observed portions of the maintenance work performed on
valve CV-8141A which isolated seal leakoff from the "A" reactor coolant
pump.
Some of the wiring which ran from a pull box to the solenoid valve
and to the limit switches of CV-8141A exhibited insulation damage
apparently from high temperature conditions.
The affected wiring was
replaced and necessary in-line splices completed.
Quality control
coverage was present to verify the acceptability of the splices and
completed heat shrink tubing.
The inspectors examined the damaged wiring
which showed brittle insulation.
The portions of the wiring which had
been located in the pull box, solenoid, and limit switches showed a
greater degree of insulation hardness than the portions which were
located in the connecting conduits.
The affected wiring did not have a
protective cable outer jacket.
The outer jacket from the associated
cable was removed at a pull box, and examination of the conductors in the
cable showed no insulation degradation.
The inspectors reviewed the work
instructions of maintenance request MR 88-4206, and the completed work
package which included an electrical termination record.
During a walkdown of cabling in areas around reactor coolant pumps "A"
and "D", the inspectors noted the identity of various cable trays and
conduits.
A review of the work package associated with MR 88-4106 and MR
88-4197 confirmed that licensee personnel had included these cable trays
and conduits in the inspection program to determine the extent and cause
of cable degradation.
Continued licensee actions to determine root cause of observed cable
degradation and implement response actions will be followed in routine
inspection activities.
No violations or deviations were identified.
5.
Containment Local Leak Rate Testing (61720)
Performence of Type B and Type C leak rate tests was controlled by
licensee periodic engineering test PET-5-2, Revision 16, titled
"Containment Local Leak Rate Testing." The inspectors reviewed PET-5-2
and verified that Data Sheet 3 included all containment isolation valves
subject to Type C testing listed in Technical Specification Table 3.6-1
and all electrical penetrations listed in FSAR Table 6.2-33.
Data Sheet
3 was used to calculate the combined leakage rate and compare the result
to the acceptance criteria of 0.6 La (maximum allowable leakage rate
specified for containment ir. the technical specifications).
At the time
of the inspection, local leak rate testing for all but a few penetrations
including the purge system exhaust and supply valves, equipment hatch,
and the 45-foot air lock had been performed.
The r.spectors verified
that the combined leakage rates including estimated (based on recent
results) leakage rates for the penetrations to be done were well below
0.6 La. Review of recent PET-5-2 data sheets showed that local leak rate
testing was performed on a yearly basis and thus met test frequency
requirements specified in the technical specifications and 10 CFR 50
Appendix J.
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The inspectors observed portions of local leak rate tests performed for
the 93-foot air lock and for the equipment hatch tool passageway.
The
tests were performed at 60 psig.
The flow and pressure instrumentation
in the LLRT test panels were verified to be within calibration.
The
inspector reviewed the applicable clearances for the tests.
The work was
performed under Radiation Work Permit RWP 88-131.
No violations or deviations were identified.
6.
Pressurizer Surge Line Elbow Replacement (62703, 93702)
In partial response to unexpected pressurizer surge line movement
discovered during the 1988 refueling outage, the licensee performed
ultrasonic examination of pipe welds on the surge line.
The results
showed relevant indications on the two welds associated with a 90-degree
elbow directly below the pressurizer.
The licensee made the decision to
remove the elbow and adjoining pieces of piping and to rework the surge
line with new components.
The process was controlled in part by Detailed
Construction Package DCP-15 Request for Design Change RDC 87-001.
The
removal of the old elbow and installation of new components was performed
with the assistance of a vendor site services crew.
The inspectors performed an inspection of the elbow replacement work
which included observations of cleanliness controls, weld rod control,
welding, plant operations interface with replacement work, review of weld
procedures, welder qualification records and material documentation.
Cleanliness controls were specified in both the licensee's maintenance
request work instructions and the vendor's procedure which was reviewed
by the Plant Review Board.
These instructions included the cleaning of
the replacement elbow and associated piping and inspection to cleanliness
criteria similar to ANSI N45.2.1 "Cleaning of Fluid System and Associated
Components for Nuclear Power Plants," Class B, conditions; an inspection
for and removal of cutting chips and other debris af ter removal of the
first section of original piping; and cleaning of the piping interior and
inspection after removal of all sections of original piping.
A pneumatic
plug was installed into the horizontal run of piping which connected to
the rest of the reactor coolant system to primarily serve as a barrier to
primary coolant.
A shielded steel plug was also inserted into the
horizontal piping and the end of the piping was covered with a plywood
disc.
On the vertical section of piping connected to the pressurizer, a
shielded steel plug was inserted and the end of the piping was covered
with a plywood disc.
The work instructions provided for tool control
during the period of time the system was open to the environment.
Upon
installation of new components, the pneumatic plug was removed prior to
fit-up of the final section of piping, and a final closeout cleanliness
inspection was performed by both licensee and vendor QC inspectors.
The inspectors reviewed material purchase orders and documentation for
weld rod, replacement elbow and piping used in the elbow replacement
work.
The suppliers were verified to be on the utility's Evaluated
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Contractors and Suppliers List.
Weld rod was stored in a locked metal
container in the vendor's trailer and in the licensee's warehouse on
shelving located off the floor in labeled boxes with controlled access
into the warehouse.
Weld rod was issued with the use of requisition
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slips and tracked on weld history records.
The inspectors observed
portions of the welding performed to install the new elbow and associated
piping and portions of the cutting to remove the original elbow and
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adjoining piping.
For welding, the vendor primarily used an automatic
pipe welding system.
The inspectors verified that a sample of the
operating parameters met the limits of the applicable weld procedure
specification.
Radiation protection and five watch coverage were
observed to be in place.
During the surge line elbow replacement, the reactor coolant system level
was maintained at slightly above the loop centerline.
A temporary
operating instruction, OIT-5-4, "Plant Operations During Surge Line
Repair," was written to control plant operation during the replacement
work and addressed establishment of a RCS overpressure path, level
control of the RCS, and decay heat removal.
The inspectors reviewed the
procedure and verified that the instructions were being followed.
Containment integrity restoration in the event of loss of decay heat
removal capability was addressed in Administrative Order A0-3-11
"Containment Access and Evacuation." The inspectors reviewed A0-3-11
worksheets which presented the restoration plans for applicable
containment penetrations and verified the completion of these plans prior
to RCS cooldown below the top of the loops.
A review was performed of a sample of the weld procedure specifications,
procedure qualification records, welder qualification instructions,
records of welder qualification tests, and specified nondestructive
examination requirements.
No substantive discrepancies were identified.
The inspectors verified that the licensee welding engineers had reviewed
the vendor procedures and had considered sensitization issues in the
review.
Four nonconformance reports had been written at the time of the
inspection.
The inspectors reviewed the reports to verify appropriate
resolutions of the nonconformances had been identified.
During the course of the inspection, it was observed that the vendor was
working seven day weeks.
The inspectors requested a sample of the time
sheets for the welding and machinist shift crews.
For the week of June
11 through June 17, 1988, members of the vendor site service crew
including welders and their immediate supervisors apparently worked in
excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a seven day period, taking into account possible
shift turnover time.
No worker showed excessive work hours relative to
the rest of the crew. Work hour limitations were controlled by
Administrative Order A0-3-1, "Shif t Complement and Work Time," which
required the use of deviation sheets if work hours limitations were to be
exceeded.
The licensee project coordinator stated that deviation sheets
had not been completed for the vendor crew.
This is an apparent failure
to follow the procedural requirements of A0-3-1 and an apparent violation
of the workiag hour limits of Technical Specification 6.2.2.g.(50-344/88-24-02).
The inspectors noted that the vendor work
proposal exrlicitly stated that the work schedule would consist of 12
hour shif ts, two shif ts per day, seven days per week.
In discussions
with the inspectors, tne licensee project coordinator indicated that a
nonconforming activity report would be initiated and A0-3-1 deviation
sheets would be promptly processed, as appropriate.
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7.
Event Follow-Up (93702, 92701)
Pressurizer Surae Line Movement
With the plant in Mode _6 and RCS temperature at approximately 80 degrees
F, engineering walkdowns of pipe whip restraints identified that the
pressurizer surge.line was in contact with the upper shim. pack of pipe
whip restraint (WR) 1.2.-
The' inspection of WR's was a result of the
licensee long. term action plan in response to pipe support problems.
identified in the 1987 refueling outage.
No pipe contact was observed
for the seven other WR's on the surge line.
The licensee determined that
approximately 5000 pounds-force was required to push the pipe off the
contact point. After the removal of two shim plates from the WR 1.2
upper shim pack, the pipe moved 3/8 of an inch in the upward direction.
The design cold gap was 9/16 of an inch.
Inspection of the "B" RCS loop
hot leg WR and steam generator upper and lower support ring revealed no
anomalies.
In response, the licensee developed an action plan to assess the cause,
significance, and corrective actions for this event.
One potential cause
which was investigated was the potential for thermal stratification in
the surge line during heatup, particularly when a bubble was being formed
in the pressurizer. At that time, a temperature difference in excess of
300 degrees F could exist between the pressurizer and the RCS hot leg.
The possibility of thermal stratification was apparently not considered
in the NSSS piping analysis.
As part of its action plan, the licensee
planned to make a determination as to the continued operability of the
surge line, make changes to piping supports as appropriate, and planned
to implement a monitoring program for temperature and deflection data for
the surge line.
Problems with RCS thermal expansion at Trojan were
encountered in 1986.
Review of these problems concluded that the RCS had
been over-restrained.
Adjustments were made to the RCS restraints and an
ongoing monitoring program was implemented.
Region V follow-up inspection of licensee actions was documented in
Inspection Report 50-344/88-25, which also dealt ~with pipe whip restraint
design issues.
Pipe Whip Restraint Desian Gaps
Licensee nonconformance report NCR 88-190 dealt with the finding that for
various pipe whip restraints, inadequate gaps existed to accommodate
calculated thermal and seismic movements without contact of the pipe to
the restraint.
Some restraints were reported to be in contact in the
cold condition.
NCR 88-190 was a generic report in that as problems were
identified with additional restraints, they were added onto NCR 88-190.
As of May 12, 1988, the plant onsite quality assurance branch copy of the
NCR showed 22 pipe whip restraints to have inadequate gaps and, of these,
eight pipe whip restraints were observed to be in contact in the cold
condition.
The pressurizer surge line pipe whip restraint 1.2 was not
among the eight or 22 restraints above, but instead was addressed in a
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separate NCR (88-204).
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The inspectors performed an inspection of the eight pipe whip restraints
which were initially reported to be in contact in the cold condition per
NCR 88-190.
In all eight cases, no gross indications of damage to the
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piping or to the restraint'itself was observed.
The as found gap data
shown in NCR 88-190 appeared to be reasonable based on visual observation
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with the following exceptions:
(1) pipe restraint 49.5 - the inspectors
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did not observe contact at the "C" shim pack; instead a slight gap of
approximately 0.020 inches was observed; (2) pipe whip restraint 5.4 -
due to close tolerances the gaps could not be observed due to the
presence of piping insulation; (3) pipe whip restraint 48.3 - this
restraint was inaccessible for close examination; the "C" shim plate
appeared to be in contact as viewed from a distance of approximately 6
feet.
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Region V follow up inspection was documented in Inspection Report
50-344/88-25.
Toward the end of the inspection period, the resident
inspectors reviewed a sample of maintenance request work packages used to
reshim pipe whip restraints to preclude contact with associated piping
during thermal and seismic transients.
The work packages provided
adequate detail and work instructions to the workers and quality control
inspectors.
A production tolerance of one eighth of an inch was
specified for shim gaps, which was verified by QC inspectors and the
results subsequently reviewed by the action engineer. The inspectors
measured shim gaps for two reworked pipe whip restraints - WR 18.3 on the
alternate charging line and WR 48.4 on the letdown line.
The results
closely matched the measurement values reported by the QC inspectors and
were within one eighth of an inch of the specified design values.
No violations or deviations were identified.
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8.
Follow-Up of Open Items (92701, 92700)
Open Item 87-30-01 (Closed) Observed Offset in Piping Supports - In
walkdowns of main feedwater piping inside containment during 1987
refueling outage, the inspectors identified two spring can hangers which
showed large offsets between the pipe clamp stanchion and the load column
attached to the spring can.
The licensee made modifications to the
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hangers (H-3 and H-6 on the "C" and "D" main feedwater lines,
respectively) to reduce the loading to acceptable values.
During the
1988 refueling outage, the licensee staff found that the adjusting rod on
hanger H-3 had sheared off from its load plate.
Nonconformance Report
NCR 88-163 was written.
The cause of the failure was reported to be
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associated either with the pinning of the hanger in preparation for
piping replacement or with hanger offset.
The hanger apparently failed
after plant shutdown for the 1988 refueling outage.
The inspectors reviewed the disposition of NCR 88-163 and discussed
corrective actions with members of the licensee staff.
The load plate on
hanger H-3 was replaced.
To reduce loadings on spring can hanger
components, the baseplate was modified on H-6 to realign the hanger.
In
addition, a 4-inch long pipe was installed between the spring can housing
and the base plate to raise the spring can on H-6.
This allowed the
adjusting rod length to be decreased.
The inspectors performed a
walkdown of all four main feedwater lines, anu noted that the adjusting
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rod for spring can hangers which support the pipe ~from below had been
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reduced to less than one inch in length.
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The inspectors reviewed licensee procedures to verify that hanger offset
was addressed.
A review of the licensee's quality control procedures
(QCP) indicated that QCP-8, Revision 19, "Hanger and Structural Steel
Inspection," did not address hanger offset and the QCP-34, Revision 2,
"Visual Examination," included instructions which would lead to the
identification of excessive hanger offset.
QCP-34 was normally used only
for inservice inspection related activities.
Installation Standard M-6,
Revision 1, which dealt with the procurement, installation, and
adjustment of piping supports, generally addressed piping offset.
It
appeared that the responsible engineer would need to include specific
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instructions in the appropriate work package to provide for the
adjustment and inspection of hanger offset.
Based on the above inspection, this item is considered closed,
LER 88-11 (Closed) Movement of New Fuel Over the Spent Fuel Pool - In
refueling activities, seven new fuel assemblies were moved with the spent
fuel pool bridge crane in the fully raised position over the spent fuel
pool.
The licensee identified this as a violation of Technical Specification 3.9.7.
The inspectors reviewed corrective actions which
included additional training for members of the refueling services
vendor.
LER 88-11 stated procedural non-compliance as the cause of the
event.
The inspectors verified that adequacy of training was discussed
in the licensee's internal event report process.
In addition, review of
the licensee's documentation indicated that the event was within the
design bases of the plant for a postulated fuel handling accident.
This
item is considered closed.
9.
Refuelina Area Material Control (71707)
An inspection was performed of material control in the refueling area
inside containment.
The licensee has implemented a quality control
procedure, QCP-16 "Refueling Area Material Control," which provided
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instructions for controlling material to prevent entry into the reactor
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coolant system.
The inspectors observed the implementation of QCP-16
measures in the refueling cavity area during fuel assembly movement.
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Mcess to the designated work area was controlled through the use of
barriers.
The material control area was designated on the QCP-16
Material Control Barrier Location Form to include the refueling cavity
and an area approximately 15 feet to each side.
General conditions were
observed to be adequate with no loose material stored in the immediate
vicinity of the refueling cavity; objects on the ma.11pulator crane were
either taped in place or attached by lanyards; and personnel in the
material control area observed requirements for securing items.
A
quality control inspector was present to control material entering the
refueling area.
The inspectors reviewed the QCP-16 notebook which
contained in part the Tool and Material Accountability Log Sheet and
found no significant discrepancies in the tool and material
accountability logs and the end of shift material accountability reports.
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The material control measures of QCP-16 were implemented as required by
the licensee's refueling procedures.
The refueling procedures also
required a thorough inspection of the refueling cavity prior to lifting
the reactor vessel head.
The inspectors performed a walkdown of the
refueling upper cavity prior to reactor vessel head lift and visually
inspected the lower cavity for the presence of foreign objects.
The
refueling floor showed good general conditions with a minimum of loose
material.
Discussions with licensee personnel and reviews of various refueling logs
indicated that on several occasions, during refueling activities, small
objects had fallen to the refueling floor from other elevations.
During
the outage, work was progressing on the steam generators and on the
205-foot elevation, as well as other areas of containment, which may have
resulted in material falling onto the refueling floor.
The inspectors
discussed the incidents with members of plant management and shared the
concern of materials potentially falling into the open reactor vessel
during fuel assembly movement, particularly in light of the reported
instances of falling objects.
On one occasion, a plastic tie-wrap
reportedly fell into the refueling cavity onto the upper internals which
was in its stand in the lower cavity.
On another occasion, five small
objects which had the appearance of weld slag reportedly fell into the
refueling cavity and landed near the ladder on the upper refueling
cavity.
The plant was reloading fuel assemblies into the core at the
time.
Plant management expressed the position that (1) controls were in place
to minimize the potential for falling objects and (2) measures would be
taken to inspect for and remove any foreign objects in the refueling
cavity.
The inspectors reviewed the controls prescribed for work
(primarily on pipe supports) performed on the 205-foot elevation inside
containment.
These consisted of the use of lanyards to secure tools and
equipment and the use of-safety belts.
The inspectors performed a
walkdown of work conditions on the 205-foot elevation and observed the
establishment of work areas through the use of scaffolding, herculite and
welding blankets which attempted to minimize the potential for dropping
objects to lower containment elevations.
The inspectors reviewed
refueling logs which indicated that the reactor vessel flange and upper
internals were inspected for foreign objects, the core was inspected
during the course of core mapping, and the refueling cavity was vacuumed
to remove potential foreign material.
No violations or deviations were identified.
10.
Followup on IE Bulletin 85-03 Item e
As requested by Item e. of Bulletin 85-3, "Motor Operated Valve Common
Mode Failures During Plant Transients Due to Improper Switch Settings,"
the licensee identified the selected safety-related valves, the valves'
maximum differential pressures and the licensee's program to assure valve
operability in their letters dated May 8 and July 15, 1986, and November
,
'
18, 1987.
Review of these responses indicated the need for additional
information which was contained in Region V letter dated March 9, 1988.
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. . .
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.
Review of the licensee's April 8, 1988, response to this request for
additional information indicated that the licensee's selection of the
applicable safety-related valves to be addressed and the valves' maximum
differential pressures met the requirements of the bulletin and that the
program to assure valve operability requested by action item e, of the
bulletin was acceptable.
The results of the inspections to verify proper implementation of this
program and the review of the final response required by action item f.
have beer. addressed in previous inspections.
This closes IE Bulletin 85-03.
11.
Exit Interview (30703)
The inspectors met with the licensee representatives denoted in paragraph
1 on June 23, 1988, and with licensee management throughout the
inspection period.
In these meetings the inspectors summarized the scope
and findings of the inspection activities.
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