ML13316C019
| ML13316C019 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/25/1989 |
| From: | Caldwell C, Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML13316C017 | List: |
| References | |
| 50-206-89-11, 50-361-89-11, 50-362-89-11, GL-83-18, GL-88-14, NUDOCS 8906090077 | |
| Download: ML13316C019 (15) | |
See also: IR 05000206/1989011
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report No.
50-206/89-11, 50-361/89-11, 50-362/89-11
Docket No.
50-206, 50-361, 50-362
License No.
Licensee:
Southern California Edison Company
P. 0. Box 800, 2244 Walnut Grove Avenue
Rosemead, California 92770
Facility Name:
San Onofre Units 1, 2, and 3
Inspection at:
San Onofre, San Clemente, California
Inspection conducted:
March 13 through 17 and
April 3 through 7, 1989.
Inspector:
,_
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C.f. Caldwe, Project Inspector
Date Signed
Approved By:
,.* o sn
1:4
P.H. Johnson, Chief,
Date Si ned .
Reactor Projects Section 3
I nspection Summary:
Inspection on March 13 -
March 17,
1989 and April 3 -
7, 1989 (Report
No. 50-206/89-11) 50-361/89-11, 50-362/89-11
Areas Inspected:
Routine unannounced regional inspection of licensee QA
program implementation, design changes and modifications, inspector followup
items, and followup on items of non-compliance.
Inspection procedures 35502,
37700, 92701, 92702, and 30703 were covered.
8 9 0
6 0 9 0 0 7 7 g 0 ,
PNu
Safety Issues Management System (SIMS) Items:
None
Results:
General Conclusions and Specific Findings:
In general, the licensee's program enhancements for quality oversight
groups were being implemented and showing a positive trend. However, the
inspector considered that additional reviews of quality oversight group
activities were necessary to verify that program enhancements were
effective.
Significant Safety Matters:
A concern was identified with the performance of reviews 'conducted in
accordance with 10 CFR 50.49 as related to the evaluation of the effect
on operation of the plant with inoperable excore neutron monitors during
post accident conditions.
Summary of Violations:
One violation of NRC requirements was identified (Paragraph 4.a) dealing
with the adequacy of the licensee's corrective actions to resolve a
10 CFR 50.49 equipment qualification discrepancy.
Open Items Summary:
One unresolved item was identified (Paragraph 4.a) with regard to lack of
the Updated Final Safety Analysis Report and the control panels to list
the source range loc power instruments as post accident monitoring
instrumentation. One followup item was identified concerning
implementation of independent safety engineering group review of
nonconformances. In addition, four followup items, one unresolved.item,
and one 10 CFR Part 21 report were closed.
DETAILS
1. Persons Contacted
Licensee Personnel
- H. E. Morgan, Station Manager
R. M. Rosenblum, Quality Assurance Manager
- R. W. Krieger, Operations Manager
V. B. Fisher, Unit 2/3 Superintendent
- M. L. Merlo, Manager, NEDO
- D. A. Herbst, Manager, Site QA
- M. P. Short, Project Manager, DBD
- L. K. Carlisle, Nuclear Engineering Supervisor
- R. D. Plappert, Compliance Supervisor
- W. W. Strom, Supervisor, Independent Safety Engineering Group
- L. D. Brevig, Supervisor, Onsite Nuclear Licensino
- R. A. Neal, Senior Engineer 2
- S. J. Foglio, Engineer 2
,#G. T. Gibson, Onsite Nuclear Licensing
- C. A. Couser, Onsite Nuclear Licensing
- R. L. Baker, Onsite Nuclear Licensing
- D. A. Werntz, Compliance Engineer
- K. D. Flynn, Engineer 1
- J. H. Martin, Engineer 1
- Denotes those attending the exit meeting of March 17, 1989.
- Denotes those attending the final exit meeting on April 10, 1989.
The inspector also contacted licensee operators, ehgineers, technicians,
and other personnel during thecourse of the inspection.
2. Evaluation Of Licensee Quality Assurance Program Implementation
(35502)
The purpose of this review was to evaluate the effectiveness of the
licensee's implementation of the quality assurance program and to
recommend, based on this evaluation, changes to the current level of
inspection effort in this area.
This effort consisted of a comprehensive
review of program surveillances and discussions with Southern California
Edison's (SCE's) quality oversight groups in order to determine if recent
changes in personnel and program direction were enhancing the
effectiveness of the quality programs to find problems and weaknesses in
SCE activities. During this inspection, the licensee's operating
experience review (OER) program and activities performed by the
independent safety assurance group (ISEG) were not reviewed. In
addition, QA audits were not reviewed during this inspection since the
licensee had not finalized any.audits for issue under the revised program
format.
Surveillances Reviewed
The inspector reviewed 20 recent surveillances that were performed of
maintenance, surveillance, and operational activities including the
fol 1
owi no:
2
-
SOS 13-89, "Technical Specification Section 4.10.A Verification"
-
SOS 29-89, "Removal Of Temporary Facility Modifications"
-
SOS 59-89, "Verify Actions In Response To 10 CFR 21 Report"
The inspector noted that most of the surveillances had findings of minimal
significance. However, several of the surveillances had substantive
findings regarding activities conducted in the plant. For example, in
SOS 89-13, personnel found that several welds that should have been
inspected during the current Unit 1 outage were not scheduled. As a
result of the finding, a corrective action request (CAR) was prepared.
Other examples were SOS 29-59 in which personnel found a support rail
that was bent out of position which would prevent a breaker from closing
and SOS 59-29 in which a procedure was found that had not been updated.
The review of surveillances indicated that the reports were more detailed
than in the past.
The inspector also discussed a recent significant surveillance findings
concerning pressurizer power operated relief valve (PORV) setpoint on
Unit 1. In particular, during a surveillance to review the stroke time
of the PORVs, QA personnel found that the stroke time for one of the
PORVs was slightly greater than the design requirement. In addition,
personnel found that the PORV setpoint for the overpressure mitigation
system and Technical Specification 3.20 had not been revised to reflect
the latest changes to the heatup and cooldown curves.
As a followup to this
finding, the licensee was evaluating the condition for safety
significance and additional corrective actions.
Principal Enhancements To Increase Quality Oversight Effectiveness
As a followup to discussions held during recent management meetings,.the
inspector discussed QA and other quality oversight program.changes with
responsible personnel.
The inspector noted that the licensee had
established or was in the process of establishing a number of program
enhancements.
Some of the more noteworthy changes were as follows:
-
Area Monitoring Program - This program was upgraded to provide the
nuclear oversight organization (all groups except for QC) with a
systematic method for direct observation of-the implementation of
established QA program requirements (e.g., housekeeping, material
conditions, and temporary modifications).
-
Quality Assurance Monitoring Report - This program was established
as an activity oriented program similar to surveillances. The
principal difference was that no paperwork reviews were done. This
report was developed to provide a method for observing plant
activities and documentation of the observations in a more efficient
manner.
-
Manpower Utilization Program -
The licensee was developing upgraded
standards and a certification process for QA personnel.
These
upgraded standards will consist of an annual requalification process
(with requirements exceeding the American National Standards Institute
3
(ANSI) standards).
The process will require a minimum number of
audits to be performed within a given period to maintain
qualification. In addition, personnel will be required to take a
requalification examination and review by a qualification board.
-
QA/QC Training - The licensee was developing a training program in
accordance with the Institute For NuclearPower Operations (INPO)
guidelines.
-
QA Audits - The licensee revised the format for audit performance so
that future audits will be more performance based (e.g., similar to
the new NRC inspection program).
-
Proactive Evaluation/Engineering Review (PEER) Program For Design
Change Documents - This program was instituted to provide the QA
organization with more effective measures to evaluate design change
documentation prior to approval.
Design Changes -
The procedures were revised to require that design
changes undergo review and comment prior to issue.
Officers' Council Weekly Progress Report On Quality Assurance -
The
licensee established this program to issue a weekly progress report
identifyinq QA actions/findings associated with recent audits,
surveillances, and other quality oversight group activities. Part
of the intent of issuing these reports was to detail issues in
such a mariner thatproper management attention could be focused to
resolution of the items.
The inspector reviewed the reports for the
weeks of March 2, 1989 through March 16, 1989.
The inspector noted
that several significant issues had been identified in the reports
for resolution. For example, one audit identified that functional
testing at end-of-discharge voltages for the Unit 1 motor operated
valve inverters and Unit 2/3 shutdown cooling valve inverters had
not been performed. As a result of this finding, the licensee
issued two nonconformance reports to evaluate the conditions.
Trending - The licensee was developing a "real time" trending
program for audits, surveillances and other quality findings.
This
program will primarily be used for trending and feedback for the
quality oversight groups.
QA Organization Employee Advancement Program -
The licensee
established this program to define a career path for employees that
details the requirements for advancement. This will include
opportunities to broaden the experience base through
inter-department cross-training.
Commitment Tracking System
The inspector selected several licensee commitments to ensure that the
licensee's system for tracking commitments was functioning properly.
The
inspector selected four commitments as follows:
4
-
Limitations placed on the six-inch vent valves in the containment
venting and purging system (Ref. letter to SCE dated November 2,
1984).
This item was identified in the San Onofre Commitment
Register (SOCR) as number 8402069/001.
-
Final report on fire protection Technical Specification changes
This item was identified in the licensing tracking list with a due
date of March 31, 1989.
-
June 17, 1988 amendment application additional information -
This
was identified in SOCR as number 8800295/001 with a due date of
March 31, 1989.
-
Response to Generic Letter 88-14 -
This was identified in SOCR as
item number 88120291/001 with a due date of May 19, 1989.
The inspector noted that these items were being tracked by the licensee
and considered that based upon this review, the licensee's tracking of
commitments was adequate.
Summary
Based upon the documents reviewed and discussions held with SCE
personnel, the inspector considered that the licensee's quality oversight
groups were headed in a positive direction. However, the inspector
considered .that it was .too early to tell if the oversight groups were
fully effective in evaluating licensee programs. The recent enhancements
provided evidence that SCE was being responsive to past NRC concerns.
The inspector considered that the true test of the effectiveness of the
quality oversight organizations would be the ability to track and resolve
recurring deficiencies, a continued reduction in the backlog of
corrective action followup items, and more QA involvement to identify
problem precursors before they reach NRC attention and concern. The
inspector will continue to review the licensee's implementation of
program upgrades to enhance quality oversight group effectiveness during
future inspections.
No violations or deviations were identified.
3.
Design, Design Changes, And Modifications
(37700)
The inspector reviewed several design changes and modifications that were
determined by the licensee to not require approval by the NRC. This was
performed in order to determine if these actions were done in conformance
with the requirements of Technical Specification (TS) 10 CFR 50.59, the
Safety Analysis Report, and the licensee's Quality Assurance Program.
The inspector reviewed the following proposed facility changes (PFCs):
-
PFC 3-87-6554.17, Revision 0, that was issued to provide an
alternate power source to the excore neutron monitoring system for
Unit 3.
-
PFC 1-88-5113.2, Revision 1, that was issued to install a 480 volt
switchgear bus-tie breaker trip on safety injection/loss of off-site
power for Unit 1.
5
PFC 2-87-6604, Revision 0, that was issued to modify the Unit 2
remote shutdown panel to bring it in compliance with the present
requirements of NUREG-0700.
PFC 2/3-87-6554.02, Revision 0, that was issued to install eight
hour emergency lights in Units 2 and 3 for Appendix R compliance.
-
PFC 1-88-5113.04, revision 0, that was issued to add an automatic trip feature for the station service transformer number three supply
breaker in Unit 1.
-
PFC 2/3-84-238, Revision 0, that was issued to add an air motor
start counter and other equipment to monitor diesel generator start
times in Units 2 and 3.
-
PFC 1-88-3410.0, Revision 2, that was issued to install level
switches at the Unit 1 refueling water storage tank.
-
PFC 1-88-3496.0, Revisiun 0, that was issued to replace the existing
Unit 1 feedwater/stem flow computer and mismatch comparator with
qualified equipment.
The inspector noted that these PFCs contained the necessary reviews and
approvals prior to issue to the station including nuclear safety group
(NSG) reviews. In addition, the inspector noted that reviews required by
10 CFR 50.59 were performed and that other considerations such as "As Low
As Reasonable Achievable (ALARA)," and fire protection were performed.
The inspector also rioted that the licensee had prepared maintenance
orders to implement the PFC/DCPs and that these maintenance orders were
properly controlled. Additional reviews of the licensee's program for
design changes and modifications will performed during future
inspections.
No violations or deviations were identified.
4.
Licensee Action On NRC Inspection Findings
(92701)
a.
(Closed) Followup Item (361/88-28-01), "Review Acceptability
Of Safety Evaluation For Gamma-Metrics Connectors"
This item identified a concern with regards to the licensee's safety
evaluation and decision to "accept-as-is" Gamma Metrics cable
connectors used on the startup log power excore neutron monitors.
On February 19, 1988, Gamma-Metrics issued a 10 CFR Part 21
notification of a potential deviation in which the connector solder
joints could leak at elevated temperatures.
The connectors were
used in Units 2 and 3 excore monitoring systems.
In this Part 21
report, licensees were requested to send spare connectors to the
vendor for testing. SCE complied with that request and sent one
connector that was located in stores. The vendor tested connectors
that were sent and revised their Part 21 report as a result of the
failures that were found.
On May 10, 1988, the vendor sent a second
letter that stated the following:
"The results of these tests
convince us that there is a significant possibility of having leaks
in cable assemblies which are installed in our customers'
6
facilities."
The vendor also stated that a leak in the cable
assembly could cause the neutron flux monitoring channel performance
to be degraded or to fail during a design basis accident.
As a result of the Part 21 report, SCE issued nonconformance report
(NCR) G-0865, Revision 0, to identify that there was a possible
problem with solder connections on Gamma-Metrics cable assemblies.
The NCR stated that the equipment was operable and that the
condition was not reportable. The interim disposition of this NCR
was to "accept as-is" per the attached safety evaluation, and the
final disposition was to rework or retest the connectors when the
qualified unit was supplied by the vendor. The inspector reviewed
the safety evaluation for the NCR that was prepared by station
technical.
It stated that there was no nonconforming condition
since the vendor tests only raise the possibility of leaks at the
solder joints.
The licensee's nuclear engineering group also
provided input (disposition and 10 CFR 50.59 evaluation) to this NCR
in a communication dated July 25, 1988. In that communication, the
engineering organization dispositioned the NCR by identifying that
the condition was not a problem since there were two redundant
trains and both would have to fail in order for the system to be
completely nonfunctional. In addition, engineering considered that
if the excore neutron monitoring system were to fail, there were
alternate monitoring means to assess core reactivity such as the
following:
1. Reactor coolant system (RCS) hot and cold leg temperature
difference.
2. Control element assembly bottom contact indicators
3. Boronometer readings.
4. RCS sample analysis.
The inspector questioned the evaluations that were performed by
station technical and the engineering organizations. In particular,
the inspector had the following concerns:
-
It did not appear that the post accident monitoring function of
the excore neutron monitors was fully considered when the
licensee evaluated them as operable.
-
The station technical safety evaluation stated that there was
no nonconforming condition despite the fact that the vendor
indicated that there was a significant possibility of having
leaks in the installed cable assemblies.
-
The licensee did not query the vendor to sufficient detail to
determine the extent of failures of the tested cable assemblies
or obtain other detailed information to aid in the operability
assessment.
-
The engineering disposition considered that the connector
problem was very significant since both trains of the excore
neutron monitors would have to fail in order for the system to
be completely nonfunctional.
This was despite the fact that
7
the vendor had identified the potential common mode failure of
the connectors.
The four alternate means of verification (mentioned above) were
not in themselves adequate to determine core reactivity during
a design basis accident. Regulatory Guide 1.97, Revision 2,
"Instrumentation For Light-Water Cooled Nuclear Power Plants To
Assess Plant And Environs Conditions During And Following An
Accident" reinforced this position by detailing the
requirements for post accident monitoring (PAM) equipment. In
particular, it defined Type A variables as those variables that
provide primary information needed to permit the control room
operating personnel to take specified manually controlled
actions for which no automatic control is provided. It also
placed these variables into three Categories to establish
further requirements for them (1 - highest, for key variables,
and 3 - lowest). Regulatory Guide 1.97 classified the excore
detectors as a Category 1 variable and Edison classified them
as a Type A variable. None of the alternate means of
verification listed by the licensee were classified as Type A
variables in the Regulatory Guide. They were identified as
verification variables only.
The licensee did not notify the control room operators that
there was the potential for inaccurate indications should
degradation of the connectors take place during a design basis
accident (DBA).
Consequently, the inspector considered that
the operators could make a judgement or take some action with
erroneous-or even a complete lack of information as to core
reactivity.
10 CFR 50.49 established requirements for the qualification of
electrical components important to safety. Paragraph 50.49(f)
requires that each item of electric equipment important to safety
must be qualified.
In addition, paragraph 50.49(j) requires that a
record of the qualification must be maintained to permit
verification that each item of electric equipment important -to
safety is qualified for its application and meets its specified
performance requirements when it is subjected to the conditions
predicted to be present when it must perform its safety function.
The NRC expanded equipment qualification policy by issuance of
several Generic Letters (GLs) including GL-88-07, "Modified
Enforcement Policy Relating to 10 CFR 50.49." This GL states that
when a potential deficiency has been identified by the NRC or
licensee in the environmental qualification of equipment, the
licensee is expected to make a prompt determination of operability
(i.e., the system or component is capable of performing its intended
design function), take immediate steps to establish a plan with a
reasonable schedule to correct the deficiency, and have written
justification for continued operation which will be available for
NRC review. In addition:
for inoperable equipment not covered by
the plant technical specifications, the licensee may continue
reactor operation if the safety function carn be accomplished by
8
other designated equipment that is qualified, or if limited
administrative controls can be used to ensure the safety function is
performed.
Based on the information reviewed, as discussed above, the inspector
considered that the licensee's NCR arid 10 CFR 50.59 evaluation did not
provide an adequate assessment or documentation in accordance with
10 CFR 50.49 requirements. In particular, the 10 CFR 50.59 evaluation
credited alternate instrumentation for accomplishing the safety
function despite the fact that none of the listed variables provides
real time core reactivity information and no administrative controls
were established (e.g., control room operators-were not notified of
the potential for these monitors to fail during post accident conditions).
This is considered a violation of NRC requirements (50-361/89-11-01).
The inspector discussed concerns with the safety evaluation for
NCR G-0865 Revision 0, with the site QA groups. In particular, the
inspector was concerned that a sufficiently critical review of the
NCR was not made (in 1988) since the disposition of the NCR was approved
by QA. The QA Manager acknowledged the inspector's concerns and
indicated that the present QA organization has been sensitized to
the need for critical reviews .of such things as NCRs.
In addition, the
inspector was informed that Edison was evaluating the need for the
independent safety engineering group (ISEG) to review the disposition
of NCRs for technical adequacy. The inspector will review the
licensee's actions on this matter which is identified as followup
item (50-361/89-11-02).
The inspector's concerns over this matter were discussed durina an
exit meeting conducted on March 17, 1989. During that meetina, SCE
committed to revising the NCR and providing a justification for
continued operation (JCO) to the NRC by March 24, 1989.
NCR G-0865, Revision 1, was sent to the NRC on March 23, 1989.
The
NCR was reviewed by the inspector and considered to be inadequate.
In particular, the inspector had the following concerns with the
revised NCR:
-
The safety evaluation and JCO relied on the same Technical
Specification instrumentation/parameters for accident
monitoring that were indicated in the original NCR.
The final disposition of the revised NCR identified that the
unqualified connectors would be repaired or replaced as
necessary prior to return to service from the Cycle 5 refueling
outage.
Due to the potential significance of this issue the inspector
conducted a followup inspection on this matter on April 3 through
April 7, 1989.
The inspector continued discussions with the
licensee in attempt to resolve this matter.
During the followup inspection the inspector identified additional
concerns:
9
The inspector reviewed the Unit 2/3 Updated Final Safety
Analysis Report (UFSAR). During that review, the inspector
noted that Table 3.11.5 and Table 7.5-2 list the post accident
monitoring instrumentation (PAMI) used in the Units.
However,
the excore neutron monitors were not listed.
A tour of the Unit 2/3 control rooms revealed that PAMI meters
are stenciled as such on the control panels. However, the
startup log power excore neutron monitors were not designated
as PAMI.
Discussions with reactor operators (ROs), senior reactor
operators (SROs), and the Unit 2/3 Superintendent revealed that
they had not been notified of the problems with the connectors
for the excore neutron monitors and that they could not be
relied upon during a DBA. In addition, the inspector also
found that the ROs and SROs were not aware that the excore
neutron monitors were PAMI.
The inspector discussed these concerns with the licensee who were
evaluating the reasons for these discrepancies.
They will be
reviewed during a future inspection as unresolved item
(50-361/89-11-03).
On April 6, 1989, the inspector continued discussions with Edison
personnel over the unqualified connectors. During this meeting, the
license indicated that they had contacted the vendor for more
detailed information concerning the failures. The results of the
discussions were that the vendor had informed Edison that all of the
connectors sent by licensees had been tested and had failed.
However, the failures were in a conservative direction (higher
indicated counts than actual) and tended to be time dependent. The
time dependency was largely contingent upon the extent of the
oricinal flaw.
The inspector discussed the concern over use of alternate monitoring
means with the licensee who stated that this instrumentation was
credited for post accident monitoring during the original safety
analysis. The inspector acknowledged this, but indicated that
Edison had committed to Regulatory Guide 1.97 since the original
safety analysis and .that the Regulatory Guide specified the excore
neutron monitors as the primary core reactivity monitoring
instrumentation. With regards to completing repairs to the
connectors during the next refueling outage, the inspector pointed
out that the next refueling outage for Unit 3 is scheduled for fall
1990 and that the licensee would not be able to determine core
reactivity (real time) if a DBA should occur.
These concerns were discussed in the followup exit meeting of April
10, 1989.
During that meeting, SCE representatives committed to
perform a reevaluation of the ability of other designated equipment
to accomplish the post-accident safety function of monitoring core
reactivity and to revise the NCR to provide for repair or replacement of
the unqualified connectors during the next outage of sufficient duration
10
after receiving a qualified unit from the vendor. Revision 2 of the
NCR and JCO were reviewed by the NRC on April 14, 1989. The licensee
indicated that compensatory actions would be established for the
operators to evaluate questionable readings from the excore neutron
monitors and borate the reactor unless a sufficient shutdown margin
could be demonstrated. In addition, the licensee committed to
inspect, test and repair, or replace the connectors during the next
outage of sufficient duration (no later than the completion of the Cycle
5 refueling outage).
The inspector considered that the evaluations in
Revision 2 of the NCR and JCO were acceptable. Also, the licensee
was aggressively pursuing correcitve actions with the vendor.
b. .(Closed) Unresolved Item (206/87-02-02), "Identification Of
Unqualified AMP Butt Splices"
This item identified a concern raised during an NRR Vendor Program
Branch inspection in February and March 1987. The concern related
to the question of whether AMP butt splices were actually used in
harsh environments.
This item was readdressed in inspection report
50-206/88-17. At that time the inspector also raised questions as
to the operability conclusion of the non-qualified splices found
and whether inspection of additional splices was necessary.
The inspector reviewed the status of this item and noted that the
licensee evaluated this condition and reported conditions found in
licensee event report (LER) 87-06, Revision 1. In this LER, Edison
indicated that a walkdown was performed to determine if there were
any unqualified splices.
During this review, the licensee noted
that the splices for both trains of the containment wide range water
level instrumentation were unqualified. The licensee subsequently
repaired the unqualified splices.
The root cause of the condition
was attributed to an erroneous deletion of termination details
on a revision of a drawing note associated with those particular
terminations. The licensee identified that they inspected
approximately 60% of the butt splices located inside the Unit 1
containment. Of those, four splices were found to be unqualified.
Since the unqualified splices could be attributed to a unique cause,
the inspector considered that the licensee's actions were
appropriate. This item is closed.
C.
(Closed) Followup Item (206/87-20-01), "Followup Of IIT Items
For Cycle 9 Refueling Outage"
This item identified a number of cycle 9 outage items that had not
been completed that were being tracked to completion. The inspector
reviewed the remaining items on the list and discussed their status
with licensee personnel. The results of this review are as follows:
(1) Electrical cable monitoring program - The licensee purchased a
system to analyze cables and a program was established and
proceduralized (Procedure SO1 XXVII-30, "Cable and Electrical
Apparatus Monitoring Program").
Discussions with the licensee
revealed that they had tested approximately 500 circuits which
roughly corresponded to approximately 1600 cables. Cables were
11
selected based on maintenance records or if their adequacy was.
questionable. The results from these tests were being
evaluated.
(2) Third auxiliary feedwater pump procedures - The licensee
indicated that the emergency operating instructions were
completed and in the review process.
Regular operating
instructions were being prepared at this time to include the
new system design. Completion of the revised operating
procedures will be performed by the next refueling. The NRC
was reviewing the new EOI guidelines for approval.
(3) Nitrogen backup for main feedwater regulating valve - The
licensee issued design change package (DCP) 3472 to install the
backup nitrogen. This was implemented during the Cycle .X
outage.
(4) Check valve wear test -
The licensee performed an assessment of
the accelerated check valve wear test results and concluded
that continuous operation will result in wear exceeding
manufacturers recommendations with one to two additional cycles
of power operation. As a result, the assessment recommended
procurement of modified replacement discs.
The licensee was
evaluating the recommendations from this review.
The inspector considered that the licensee's actions (and proposed
actions) on these items were appropriate. Therefore, this followup
item is closed.
d.
(Closed) Part 21 Report (50-206/88-06-P), "Defective Replated Clevite
Bearings Used In General Motors Diesel Generators"
This Part 21 report identified a defect in Clevite upper connecting
rod bearings used in the emergency diesel generators manufactured by
General Motors Electro Motive Division.
This item is not applicable to Unit 1 since Transamerica Delaval
Inc. diesel generators are used and the licensee did not install any
Clevite bearings in the Unit 1 D/G. This item was closed for Units
2 and 3 in inspection report 50-361/88-28; 50-362/88-27.
e.
(Closed) Followup Item (206/88-02-01), "Annual 50.59 Report
Review For All Units"
This item concerned the fact that the licensee's annual facility
modification evaluation report dealt mainly with modifications that
had been funded rather than completed. The concern was, if the
decision was made later to not install the modification, how would
this be handled.by the licensee and what effect did it have on
subsequent submittals (i.e., Safety Analysis Report updates).
The inspector discussed this concern with Edison personnel who
indicated that the annual report was not used by SCE other to inform
the NRC of intended modifications. Since it was not used to track
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the status of these modifications or used for updating the Safety
Analysis Report, the licensee did not consider that it would be
confusing to them should a modification not be installed. They also
considered that, if a proposed modification were altered during
installation, it would be identified to the NRC during the next
annual update. Based upon this conversation, the inspector
considered reporting changes in this manner was satisfactory. This
item is closed.
f.
(Closed) Followup Item (206/87-12-01), "Followup On Licensee's
Justification For Amount Of Time To Implement A Design Change"
This item identified a concern with regard to theamount of time to
implement a design change for the containment spray low flow
annunciator. In particular, the design change was prepared in 1979,
but had not been implemented for 9 years.
The inspector discussed this concern with the licensee and learned
that most of the work to install a containment spray low flow
annunciator had been completed during this outage. The balance of
the work was considered to be workable with the plant on-line
and is currently planned for completion after return to service.
The licensee indicated that the long period of time to implement the
design change-was a result of a transfer in methodology for making
design changes that occurred in 1979. Edison indicated that in the
old process they could not tell what had been implemented and what
hac not.
The current program requires that the drawings be revised
when the work is completed. In addition, the DCP process requires
as-built verification of the elementary drawings. The licensee
indicated that they were not aware of any other problems such as
this. However, walkdowns of the P&IDs to compare with as-built
conditions was in process as part of another program. The inspector
considered that the licensee's actions were appropriate and that the
as-built verification process should be adequate to find additional
discrepancies. Therefore, this item is closed.
g.
(Closed) (206/TI-15-91), "Licensee's Actions In Accordance With
This inspector reviewed the licensee's modifications to the reactor
trip breakers per the requirements of Temporary Instruction 2515/91
and documented the results in inspection report 50-206/88-12;
50-361/88-09; and 50-362/88-09. In that report, the inspector
identified that this issue was closed for Units 2 and 3, but it
would remain open for Unit 1 since there were no provisions for
on-line testing of the Unit 1 Reactor Trip breakers inspectors.
Since that inspection, this item was resolved in a letter to the
licensee dated September 16, 1988 in which NRR agreed that on-line
testing of these breakers was not required. Therefore, this item is
closed.
One violation was identified in Paragraph 4.a.
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5. Exit Meeting
(30703)
On March 17, and on April 7, 1989, exit meetings were held with the
licensee representatives identified in paragraph 1. The inspector
summarized the inspection scope and findings as described in this report.
During these meetings, SCE representatives acknowledged the violation
identified in Paragraph 4.a. In addition, the licensee committed to
performing a reevaluation of the operability of the excore neutron
-monitors and providing justifications for continued operation for NRC
review as discussed in Paragraph 4.a. These commitments were
satisfactorily met by the licensee.
The licensee did not identify as proprietary any of the materials
reviewed or discussed during this inspection.