ML20235G254
| ML20235G254 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 06/23/1987 |
| From: | Rebecca Barr, Mendonca M, Suh G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20235G225 | List: |
| References | |
| 50-344-87-18, NUDOCS 8707140178 | |
| Download: ML20235G254 (13) | |
See also: IR 05000344/1987018
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U.S. NUCLEAR REGULATORY COMMISSION
REGION Y
Report No. 50-344/87-18
Docket No. 50-344-
License No. NPF-1
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Licensee: .
Portland _ General Electric Company
4
121.S. W. Salmon Street
,
Portland, Oregon 97204
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Facility Name: Trojan
Inspection at:
Rainier, Oregon
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Inspection conducted: April 12 - May 23, 1987
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Inspectors:
1'NN
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4/z>/P7
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R. C. Barr
Date Signed
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Senior Resident Inspector
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K%
m .4 /< m
G. Y. Suh
Date Signed
Resident Inspector
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m_
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Approved.By:
M. M. Mendonca, Chief
Date Signed
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Reactor Projects Section 1
Summary:
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Inspection on April 12 - May 23, 1987 (Report 50-344/87-18)
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Areas Inspected: Routine inspection of operational safety verification,
maintenance, surveillance, refueling activities, and follow-up on previously
identified items.
Inspection procedures 30703, 60710, 61726, 62703, 71707,
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71710, 90712, 92700, 92701 and 93702 were used as guidance during the conduct
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of the inspection.
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Results:
Of the areas inspected, apparent violations involving failure to obtain a
deviation from Trojan operating procedures, Technical Specification 6.8.1,
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. and failure to establish adequate controls using QC hold tags or danger and
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caution tags were identified.
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870714017G B70623
ADOCK 05000344
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DETAILS
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1.
Persons Contacted
- C. A. Olmstead, Plant General Manager
R. C. Jarman, Manager, Nuclear Quality Assurance Department
R. P. Schmitt, Manager, Operations and Maintenance
D. R. Keuter, Manager, Technical Services
J. K..Aldersebaes,' Manager, Plant Modifications
J. D. Reid, Manager, Plant Services
R. E. Susee Operations Supervisor
R. L. Russell, Assistant Operations Supervisor
D. W. Swan, Maintenance Supervisor
R. A. Reinart, Instrument and Control Supervisor
T. O. Meek, Radiation Protection Supervisor
R. W.-Ritschard, Security Supervisor
R. E. Fowler Manager, Mechanical Engineering
C. M.- Dieterle, Supervising Engineer, Nuclear Plant Engineering
G. A. Zimmerman, Manager, Nuclear Regulation
- D. K. Nordstrom, Engineer, Nuclear Regulation
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.
Operational Safety Verification
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.
On a daily basis, 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 all pertinent information was relayed to the oncoming
shift personnel.
During each week, the inspectors toured the accessible areas of the
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facility to observe the following items:
a.
General plant and equipment conditions,
b.
Maintenance requests and repairs.
c.
Fire hazards and fire fighting equipment.
d.
Ignition sources and flammable material control.
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Conduct of activities in accordance with the licensee's
administrative controls and approved procedures,
f.
Interiors of electrical and control panels.
g.
Implementation of the licensee's physical security plan.
h.
Radiation protection controls.
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Plant housekeeping and cleanliness.
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Radioactive waste systems.
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Proper storage of compressed gas bottles.
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The licensee's equipment clearance control was examined weekly by the
inspectors to determine that the licensee complied with technical
specification limiting conditions for operation with respect to removal
of equipment from service. Active clearances were spot-checked to ensure
that issuance was consistent with plant status and maintenance
evolutions.
During 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 aligned with the work activities
involved.
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The inspectors examined the licensee's nonconformance reports (NCR) to
confirm that deficiencies were identified and tracked by the system.
Identified t.anconformances were being tracked and followed to the
completion of corrective action.
Logs of jumpers, bypasses, caution, and test tags were examined by the
inspectors.
Implementation of radiation protection controls was verified
by observing portions. of area surveys being performed, when possible, and
by examining radiation work permits currently in effect to see that
prescribed clothing and instrumentation were available and used.
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Radiation protection instruments were also examined to verify operability
and calibration ste.tus.
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.
ESF systems
verified operable during this inspection period included the "A" train of
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the Emergency Power Generation System.
No violations or deviations were identified.
3.
Maintenance
Motor Operated Yalve (M0V) Maintenance
The inspector observed selected portions of the licensee's preventive
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maintenance efforts on MOVs. The licensee initiated this program on
approximately 122 out of 138 Environmentally Qualified MOVs for the
current refueling outage.
In this program the licensee found several
MOVs with frayed electrical lead wires. The licensee initiated
corrective maintenance activities of repotting and resleeving the M0V
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lead wire. The inspector observed this work which was conducted in
accordance with Field Change Notices #5 for RDC 86-033 DCP2. The
involved engineers and maintenance personnel seemed knowledgeable and
qualified for the maintenance activity.
Other discrepancies obrerved by the licensee included one brake
application problem, one loose stake in a stem nut, and discolored or
separated grease. The c:tuse of the brake application problem has not
been identified, but the brake has been replaced. The loose stem nut has
been attributed to inadequate stake installation. The discolored or
separated grease was determined to be an aging effect, and analysis of
the grease found it would still maintain its lubricating qualities.
The
inspector observed the discolored grease and reviewed licensee analyses
based on similar findings at the LaSalle plant in 1985.
Additionally, the licensee removed a large number of terminal blocks from
MOV motors as unnecessary and connected lead wires directly to motor
studs. The licensee also replaced a small number of cracked limit and
torque switches.
Finally, the licensee found a number of wire nuts and
taped connections for the valves and replaced these connections. The
inspector observed selected portions of the corrective maintenance
activities on these topics.
Problems observed at another plant were also discussed with the licensee.
Specifically, M0V over thrust conditions and valve internal damage, and
undersized power cables on MOV operators and pickup / dropout voltage
problems. The licensee's recent M0 VATS testing identified some minor
problems in valve thrust conditions in relation to possible motor
overcurrent conditions which have been corrected, but no indication of
valve internal damage has been found. Also, the licensee has measured
and is measuring motor voltage conditions and found no problem in
relation to cable size or voltage condition.
The licensee is currently evaluating the significance of these findings,
but preliminarily has determined that the findings did not impact valve
operability (0 pen item 87-18-01).
Auxiliary Feedwater Pump Diesel Engine
The inspectors observed maintenance work being performed on the diesel
engine of the diesel-driven auxiliary feedwater pump. The work was
controlled by Maintenance Request (MR) 87-1485 which was prepared in
response to Event Report 86-124. ER 86-124 reported the failure of
gasketed compression-type fittings in the copper lubricant lines during
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routine testing of the diesel engine. MR 87-1485 provided for the
replacement of all lube oil fittings and copper lines with compression
type fittings and stainless steel lines. The inspectors verified that
required administrative approvals and tagouts were in place, that the
work instructions were adequate to control the activity, and that quality
control hold points were being observed. The work was being performed
with continuous quality control coverage during disassembly and
reassembly of the diesel engine,
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4.
Surveillance
The surveillance testing of safety-related systems was witnessed by the
inspectors. Observations by the inspectors ir.cludea . verification that
proper procedures were used, test instrumentation was calibrated and that
the system or component being tested was properly removed from service if
required by the test procedure.
Following completion of the surveillance
tests, the inspectors verified that the test results met the appropriate
acceptance criteria. Any necessary corrective maintenance was completed
during the conduct of the tests in accordance with approved maintenance
request (s). Surveillance tests witnessed during this period included:
P0T-10-2 DA, " Fire Protection System, Motor Driven Fire Pump Performance."
No violations or deviations were identified.
5.
Refueling Activities
During the current refueling outage, the inspectors conducted an
inspection of activities to determine whether pre-refueling activities
specified in the Technical Specifications have been completed and whether
refueling activities are being controlled and conducted as required by
Technical Specifications and approved procedures. The licensee primarily
uses two procedures to control the conduct of refueling operations and
for establishing and maintaining control of plant conditions.
Fuel
Handling Procedure (FHP) 5-2, titled " Refueling Procedure," serves as the
controlling document governing the refueling operation.
FHP 5-2 includes
checkoff lists for the performance of surveillance testing required to be
performed as prerequisites for refueling, for checkout of tools and
equipment for proper functioning, and for control of the actual fuel
handling operations. General Operating Instruction (G0I) 11, titled
" Plant Operation for Refueling," is used to establish compliance with
surveillance testing requirements prior to the start of and during
performance of core alterations. G01-11 uses a checkoff list which is
maintained by the control room operator with completed copies of the
checkoff list to be kept with the Fuel Handling Senior Reactor Operator's
record copy of FHP 5-2.
The inspectors performed various walkdown inspections of the refueling
cavity, fuel transfer canal, and spent fuel pool areas during the
performance of equipment checkout and of fuel handling operations. The
inspectors verified that the refueling crew in containment and at the
spent fuel pool had copies of the current revision of FHP 5-2, that
direct communication was being maintained between the control room and
the refueling crew, that an engineering advisor was present in the
control room during the performance of fuel handling operations, and that
tool and item accountability was being maintained by a quality control
representative who maintained a tool accountability log.
The record copy of FHP 5-2 kept by the Fuel Handling Senior Reactor
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Operator (SRO) was reviewed. The inspectors verified that quality
assurance and quality control hold points were being observed and signed
off and that completed copies of the G01-11 checkoff list were being kept
in the Fuel Handling SR0's Fuel Handling Procedure binder. The
inspectors reviewed a sampling of applicable surveillance t! sting data
sheets and the control room log to confirm that technical s) edification
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surveillance tests were performed in a timely manner.
In particular, it
was verified that containment integrity was established within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />
prior to the start of core alterations and that channel functional
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testing of the source range neutron flux monitors was performed within 8
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hours prior to the start of core alterations. The inspectors also
reviewed the " Refueling Periodic Surveillance Checkoff List" of G01-11 to
confirm that required surveillance tests were being performed during Mode
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The inspector confirmed that the licensee had conducted a 10 CFR 50.59
safety evaluation of the reload core which was documented in License
Change Request 87-10. The safety evaluation discussed changes to the
licensee's facility as described in the updated FSAR which included
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reevaluation of the rod ejection accident for beginning of cycle
conditions and revised margins and limits for the shutdown margin
analysis. The licensee's evaluation concluded that these changes did not
involve unreviewed safety questions.
No violations or deviations were identified.
6.
Followup on Open Items
Followup Item 86-31-01(Closed): Based on review of a similar occurrence
at another plant, the inspectors had identified to the licensee an
apparent discrepancy in the valve stroke time for the pressuri7er power
operated relief valves (PORVs) used in the overpressure mitigt. tion
system. The licensee's Technical Specifications and inservice test
procedures did not specify acceptance criteria for PORV strohe times.
The safety evaluation for the technical specification ammendnent that
deals with the overpressure mitigation system assumed a PORV opening time
of 0.6 seconds. A review of recent PORV inservice tests showed opening
times which ranged from 2.8 to 9.2 seconds.
In response to this apparent discrepancy, the licensee prepared
Nonconformance Report 86-215 which declared the overpressure mitigation
system to be inoperable and initiated corrective action and an evaluation
of contributing causes as documented in Event Report 86-119. The cause
of this event was determined to be lack of documentation and cognizance
of the opening time requirements for the PORVs. A reanalysis was
performed on the impact of the longer PORY opening times on the ability
of the overpressure mitigation system to perform its design function.
The overpressure mitigation system provides protection against pressure
transients during cold shutdown, heat up, and cooldown operations in
order to minimize the potential for impairing reactor vessel integrity
when operating at or near the vessel ductility limits (as specified in 10 CFR Part 50 Appendir G).
The inspector reviewed Event Report 86-119 and the reanalysis of the
overpressure mitigation system with the longer PORV opening times. The
reanalysis supported the conclusion that overpressure mitigation system
met technical specification operability requirements with a PORV opening
time of 10 seconds. The reanalysis demonstrated that either one of the
two PORVs had adequate relieving capability to protect the reactor
coolant system from overpressurization when the transient is limited to
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either the start of an idle reactor coolant pump with the secondary water
temperature of the steam generator 50 degrees Fahrenheit above the
. reactor coolant system cold leg temperature or the mass addition from
sircultaneous operation of one safety injection pump, two centrifugal
charging pumps, and one positive displacement pump injecting into a water
solid system.
Based on the results of this reanalysis, the overpressure
mitigation system'was declared to be operable. The licensee's other
corrective actions included plans to prepare a Design Basis Document for
the overpressure mitigation system per Commitment Tracking List item
20115 and the addition of PORV stroke time requirements in the Inservice
Testing Program. The ins
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Periodic Operating Test (pectors verified that the current revision of
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POT) 1-5, which provides instructions for
full-stroke inservice exercising of the pressurizer PORVs, block valves,
and reactor vessel head vent valves, specifies PORV stroke times
consistent with the assumptions made in the reanalysis. This item is
considered to be closed.
LER 87-06 (0 pen): The licensee reported in Revision 0 of LER 87-06 that
flood relief louvers in the west wall of the turbine building were
inadequately designed to prevent flooding of safety-related equipment
located on the 45 foot level of the turbine building. The safety-related
equipment include the emergency diesel generators, auxiliary feedwater
puns, and the remote shutdown panel. The turbine building internal
flooding design bases as stated in the updated FSAR assumes a maximum
flood rate of 500,000 gpm from a failure in the circulating water system
piping. Safety-related equipment are located in rooms protected against
flooding up to elevation 47 feet by flood dikes. At the time of
discovery, the flood relief louvers and a large portion of the turbine
building wert wall had been removed in preparation for condenser tube
changeout performcd during the current refueling outage. This would have
provided an adequate flood relief capacity in the event of a failure in
the circulating water system piping.
The inspec', ors held discussions with licensee representatives and
reviewed tne licensee's documentation, including Nonconforming Activity
Report (NLAR) P87-011M, Event Report 87-018, and Temporary Modification
(TM)87-045.
The determination that the flood relief louvers were
inadequately designed resulted from the evaluation of NCAR P87-001M,
which reported that the flood relief louvers had been temporarily
replaced by plywood and sheet metal in the process of removing a portion
of the turbine building west wall for condenser tube changeout work. The
evaluation concluded that the plywood and sheet metal arrangement
provided flood relief capacity equivalent to the louvers, but also
identified design errors in the original plant design of the flood relief
louvers. The licensee's corrective actions include the following. The
flood relief louvers will be redesigned to provide less flow resistance
with the area occupied by the louvers to remain open in the interim.
Safety related equipment located on 45 foot elevation of the turbine
buildir g will be protected to elevation 48 feet by raising the flood
dikes per TM 87-045. The flood dike extensions are currently being
installed. The licensee's calculations indicate that the turbine
building would be able to withstand a 500,000 gpm flood rate and prevent
flooding of safety-related equipment with flood dikes raised to 48 feet
and redesigned flood louvers. The licensee will also submit revised
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turbine building internal flooding design bases to the NRC. Based on
discussions with the NRR project manager, t.he inspectors understand that
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NRR plans to review the licensee's revised design bases and applicable
revisions to the updated FSAR. This item will remain open pending
completion of licensee corrective actions.
Information Notice No. 86-53 and LER 87-09 (0 pen):
IE Information Notice No. 86-53, " Improper Installation of Heat Shrinkable Tubing", dated June
26, 1986, alerted licensees to a potentially generic safety problem
involving improper installation of heat shrinkable tubing manufactured by
Raychem over electrical splices and terminations. As discussed in the
information notice, if actual installation configurations do not conform
to the manufacturer's installation instructions, the status of the
equipment qualification of safety-related electrical equipment would be
indeterminate and the licensee would need to take appropriate actions to
establish compliance regarding system operability.
Upon receipt of IN 86-53, the licensee initiated its preparations for inspecting Raychem
splices on environmentally qualified safety-related equipment. The
preparations included compiling a list of all applicable equipment which
were documented as having Raychem heat shrinkable tubing splices,
purchasing Raychem splice material to replace those splices destroyed
during the inspection process and splices determined to need rework as a
result of the inspections, preparing inspection plans and training
licensee personnel on inspection and installation mettods. Review of the
listing of applicable Raychem splices to be inspected showed that most of
the splices were located inside containment. Actual inspection was
initiated during the current refueling outage.
As reported in LER 87-09, the licensee's inspections revealed a number of
Raychem splices, including splices on electrical connections for the
solenoid valves which control the pressurizer power operated relief valve
and splices on containment penetration power and control valves, which
were not installed in accordance with Raychem installation instructions.
Most of the splices had been installed during original plant construction.
Installation defects included excessive bending of the Raychem tubing,
the use of improperly sized tubing, and the lack of adequate insulation
tubing length. The inspectors discussed the inspection program with
licensee representatives, inspected two cases of Raychem splices which
failed the inspection criteria, and reviewed Maintenance Request (MR)
87-1758 which controls the inspection and rework of applicable Raychem
splices. Additional NRC inspection of the licensee's program is further
discussed in Inspection Report 50-344/87-13.
The licensee's corrective actions were to expand the scope of the
inspection program from its initial sampling basis to a complete 100%
inspection of environmentally qualified Raychem splices inside and
outside the containment, to rework all splices which fail to meet the
inspection criteria, and to rework all Raychem splices on containment
penetration cables where a high failure rate on inspection was
experienced. Exceptions were made to the above which included cases
where the applicable splices would be subject to rework in concurrent or
near-term design change packages and cases where recent maintenance work
had been performed with documentation which showed the Raychem
installation instructions had been followed.
The Raychem splice
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inspection and rework program is currently nearing completion. The
inspectors will. complete review of this item upon completion of licensee
corrective actions.
7.
Event Follow-Up
A.
Fill Line Failures of the 'A' Cold Leg Safety Injection Accumulator
On May 12, 1987, while attempting to transfer borated water (sluice)
from the 'A' to 'D' Cold Leg Safety Injection (SI) Accumulator, the
fill line of the 'A' Accumulator failed near the fill line to vessel
nozzle weld. The reactor was in the refueling mode and the
accumulators were not required by technical specification. The
licensee generated a ' ROUTINE' event repcrt and assigned maintenance
as the event evaluator. Additionally, a Nonconformance Report (NCR)
was generated by Quality Assurance and assigned to maintenance for
dispositioning. From May 12, 1987, maintenance reviewed the event
for root cause. On May 21, 1987, maintenance concluded engineering
expertise was needed to determine the root cause of the fill line
failure and transferred responsibility of dispositioning the NCR to
the Nuclear Plant Engineering Department (NPED). Preliminary NPED
determination based on a metallurgical analysis indicated the
failure was caused by low cycle-high load fatigue failure. The
failure was initially believed to be caused by individuals walking
on the fill line. [ Subsequent calculations confirmed that
-individuals walking on the fill line could not have caused the
failure.]
On May 23, 1987, following the weld repair of the fill line and a
hydrostatic test of the weld, the licensee again attempted to
transferboratedwater(sluice)fromthe'A'to'D'SIaccumulators
via the fill lines. On the first attempt to sluice in this manner,
the operator stopped sluicing when a great deal of noise was
observed by personnel inside containment. After discussions with
operators that had previously been involved in such sluicing
operations, it was determined that the noise was normal for sluicing
operations, ar.d the sluicing restarted. Operators stopped sluicing
a second time when plant personnel notified the control room of a
great deal of noise around the accumulator. Again operators stopped
sluicing, but restarted once they convinced themselves that it was
not unusual. This sluicing resulted in failure of the fill line of
the ' A' SI accumulator again with the reactor in the refueling mode.
An ' URGENT' event report was generated, the Performance
Monitoring / Event Analysis Group was immediately activated and the
system was quarantined.
Licensee's procedure 01 5-2 required that
transfer of borated water between SI accumulators be accomplished
using the accumulator sample lines vice the fill lines. A deviation
to the procedure, as required by Plant Procedure A0 4-4, was not
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obtained prior to attempting transfer between the SI accumulators
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using the fill line on May 12 or May 23, 1987. This is an apparent
violation (50-344/87-18-02). Also, the fact that licensee failed
to establish hold tags so that the fill line would not be utilized
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prior to resolution of the problem, is considered an apparent
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violation (50-344/87-18-05). Based on the inspection of these
events, licensee personnel should have known from training and
procedure use that these actions were violations and avoided the
violation.
Further, the fact, that operators ignored several warnings of
unusual noise and the fact that this fill line had previously
failed, is a significant concern.
In a subsequent management
meeting on June 5,1987, the Regional Administrator emphasized that
this fact indicates that operators performance not only was in
violation of procedures but also potentially inconsistent with the
basic fundamental of running a nuclear power plant. Specifically,
even though operators were fully aware of abnormal noise and system
vibration that occurred for each time they sluiced in this manner,
they continued the traasfer to a second failure of the fill line.
The licensee's event analysis is expected to be complete by June 5,
1987. Additionally, the licensee has developed a detailed and
comprehensive plan of action to determine the failure mechanism,
repair the failure and prevent further event recurrence.
B.
'B' Feedwater Line Seismic Restraint Damage
On May 19, 1987, the licensee identified damage to the 'B' feedwater
line seismic restraint EBB-3-1-SR8. The knee brace anchor bolts had
pulled out of their concrete foundation. Since the damage of the
seismic restraint was identified, the licensee had developed an
action plan to determine the root cause, redesign the system
support, evaluate other potential system damage and repair the
support. On a daily basis the licensee is updating Regional
Management of their findings. The licensee has categorized the
understanding, resolution and repair of the feedwater support system
as a ready for startup item.
(0 pen,87-18-03)
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C.
Main Steam Line Wall Thickness
art of their program to identify
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the licensee, as p(wall thinning), discovered the
On May 15, 1987,
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piping that is undergoing erosion
pipe spool piece containing the steam line venturi on the 'B'
steam
line had a wall thickness of .78 inches which is less than the
minimum required thickness of .858.
Subsequent examination of the
A, C and D main steam lines also indicated a wall thickness less
than the minimum required. The licensee has developed an action
plan to determine the cause of the less than minimum wall thickness,
evaluate its impact to continued operation and determine why the
wall thinning was not recognized as a result of previous
measurements. The licensee on a daily basis is updating Regional
Management of their findings. The licensee has categorized the
resolution of this issue as a ready for startup item.
(0 pen,
87-18-04)
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Shutdown Cooling During hid-Loop Operation
On May 1,1987, a Regiorh1 based senior reactor engineer and the resident
, inspector met with the Mcensee to review the April 10, 1987, extended
7lossofResidualHeatRai. oval (RHR)coolingduringmid-loopReactor
Coolant System (RCS) operation that occurred at Diablo Canyon Unit 2.
During the meeting the event root causes, the potential. vulnerability of
the Tro,ian fccility to a like event and design chr.nges that had already
been implemented by the licensee were discussed. The licensee stated
previous difficulties which they had experienced during mid-loop RCS-RHR
operation caused them to change level indication design and operating
procedures. Since changing the level indicating system design, changing
operating procedures and training operators to the challenges of mid-loop
operation no further losses or near miss losses of RHR cooling while in
mid-loop RCS operation have been experienced. Additionally, the licensee
acknowledged the need to be able to rapidly reestablish containment
integrity and be aware of actions required to establish containment
integrity during mid-loop RCS operation. Since the meeting, the licensee-
has operated in mid-loop on RHR cooling with no difficulty. The licensee
has'since also received NRC IN 87-23: Loss of Decay Heat Removal During
Low Reactor Coolant Level Operation and placed it on their commitment
tracking log.
9.
Allegation Follow g
a.
A11egath Tracking Syntem No. RV-86-A-0097
Characterization
Potentially fraudulent Laterial supplied by Familian Northwest may
have been installed in safety-related o- important to safety
applications.
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Implied Significance to Design, Construction or Operation
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The use of potentially fraudulent material of an indeterminate
quality could result in an unanalyzed safety condition.
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Assessment of Safety Significance
The licensee was reinformed that potentially fraudulent material
could have been installed in plant applications by the Office of
Investigations in November 1986. The licensee initiated an
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investigation of procurement and maintenance records to identify
installed Familian Northwest supplied components. This
investigation identified Familian Northwest supplied products as
follows:
(1) stainless steel piping and flanges for the gaseous
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radwaste system; (2) stainless steel piping and fittings for the
gaseous radwaste return to containment line; (3) stainless steel
piping for containment hydrogen sampling system; (4) wrapped piping
and fittings for the service water strainer to the discharge and
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isolation structures; and (5) carbon steel flanges used for Reactor
Vessel Level Instrumentation System (RVLIS) support.
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11
.
The licensee's Nuclear Plant Engineering Department recently
completed an analysis of the identified Familian Northwest supplied
components (Memorandum ANR-491-87M dated April 27,1987). This
analysis assumed that the components in question failed in all
applications, except for the RVLIS support flanges.
For the
components assumed to fail, the analyses concluded that the
potentially fraudulent material would not have an adverse impact on
FSAR safety analyses or an unreviewed safety question. The
analyses, for the RVLIS support flange, verified by visual
examination and engineering judgement, that the material would not
fail in.its application.
An NRC inspector had independently observed selected portions of the
licensee's investigation of procurement and maintenance records to
identify Familian Northwest supplied components and had observed
samplingofthreeofthecomponents(InspectionReport86-47). The
three samples were taken from components on the gaseous radwaste
system. All samples were analyzed by an independent laboratory for
the NRC and were verified to be the material specified for the
component. During the current outage, the licensee plans to remove
the Familian Northwest supplied stainless steel pipe from the
containment hydrogen supply system. The licensee plans to have the
removed piping material tested, and provide the results to the NRC.
When the questioned components were originally identified, the
Region V management and inspection personnel independently
determined that failure of the components would not increase the
probability of an accident or result in an accident that was more
severe than the FSAR accident analysis. The staff verified that
failure of the Familian Northwest supplied components on the
radwaste system would not result in any gaseous radioactivity
release beyond the limiting analysis in FSAR section 15.7.
The
staff also concluded that failure of the components in the
containment hydrogen sampling system would still allow sampling of
containment hydrogen from the other train and would not effect
containment isolation capability as required by the FSAR for this
system.
For the service water line to the discharge structure, the
(
staff determined that the failure of the Familian Northwest supplied
components would not result in any additional loss of service water
flow beyond that already assumed in the FSAR.
Finally, the staff
concluded that the RVLIS support flanges would not affect FSAR
analyzed accidents in that the RVLIS is an indicating system which
has specified procedural and/or equipment ahernates in case of its
failure.
The inspector reviewed the licensee's engineering evaluation of the
Familian Northwest supplied components. The inspector's review of
the licensee's evaluation concurred in the licensee's conclusion
that there was no safety concern in relation to the Familian
Northwest supplied components. As previously discussed, this
evaluation was verified by the independent staff review which
assumed component failure for the Familian Northwest products.
,
.
.
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12
,
.
Staff Position
The staff determined that (1) based on samples taken, the Familian
Northwest supplied materials were probably as specified by design
and (2) that failure of the Familian Northwast supplied components
would not result in an increased probability or more severe
consequences than already analyzed in the FSAR.
Action
None
10. Exit Interview
The inspectors met with the licensee representatives denoted in paragraph
1 on June 4, 1987, and summarized the scope and findings of the
inspection activities.