ML20132E900
| ML20132E900 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/09/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20132E882 | List: |
| References | |
| 50-373-96-16, 50-374-96-16, NUDOCS 9612240022 | |
| Download: ML20132E900 (15) | |
See also: IR 05000373/1996016
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION lli j
Docket Nos.: 50-373, 50-374
Report Nos.: 50-373/96016, 50-374/96016 l
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Licensee: Commonwealth Edison Company
Facility: LaSalle County Station, Units 1 and 2
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Location: 2601 North 21st Road I
Marseilles, IL 61341 !
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Dates: October 1 - October 8,1996 l
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Inspectors: Frederick D. Brown, Project Engineer .
Andrew Dunlop, Reactor Engineer l
Dell R. McNeil, Reactor Engineer
Approved by: B. L. Jorgensen, Acting Chief
Reactor Projects, Branch 5 ]
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9612240022 961209
G ADOCK 05000373
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EXECUTIVE SUMMARY
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LaSalle County Station, Units 1 and 2
NRC Inspection Report 50-373/96007(DRP); 50-374/96007(DRP)
This inspection report includes aspects of licensee operations and engineering. The report l
covers a special inspection performed by regional inspectors in response to licensee l
identified problems with Technical Specification (TS) compliance when an Emergency
Diesel Generator (EDG) was taken out of service (OOS), and with air operated valves
(AOVs) which provided a Primary Containment isolation system (PCIS) function. The ,
purpose of this inspection was to assess the licensee's response to each specific issue,
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and to determine whether the licensee's staff had exhibited a proper safety consc,ousness
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as each issue was identified.
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Plant Ooerations
e The licensee identified that irradiated fuel movement was conducted while
Containment Purge and Ventilation system valves were in a condition outside the
applicable TS requirements. This condition was the result of inadequate attention
to detail and poor review of an OOS on the part of licensed SROs. When the
condition was identified, the on-duty operators misinterpreted the TS time clock. )
The licensee effectively communicated the need for SRO TS interpretation l
improvement to the staff. The inspectors identified a lack of formal documentation '
of a policy for OOSs associated with work on EDGs. No indications of a lack of I
proper safety consciousness were observed. A non-cited violation was identified j
(Section 01.2).
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- Four channels of the Intermediate Range Monitoring (IRM) trip function were
inadvertently removed from service under an OOS for the Average Power Range I
Monitoring (APRM) trip function. Compliance with the Reactor Protection System )
(RPS) TS was achieved fortuitously. The poor performance on the part of the SROs I
involved with this event is similar in nature to that described in Section 01.2, and
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previously identified in inspection Report 50-373/374-96007 (Section 01.3).
Enaineerina
e in March 1996, the licensee identified a problem with the adjustment of spring
preloading for AOVs with a PCIS closure function. In September 1996, the
licensee's preliminary determination, based upon conservative calculations, was I
that some AOVs would not close under their design basis dynamic loads. The
licensee intended to perform as-found spring preload tests to determine whether the I
valves would have actually generated the required design basis closing and seating ,
forces. The inspectors determined that the problem had not been formally I
documented by a PlF and had not received a formal operability assessment in
March or April 1996 as required by plant procedures. This was characterized as a
violation of 10 CFR 50, Appendix B, Criterion V (Section E1.1).
- The inspectors identified a potential weakness in the licensee's implementation of
NRC guidance on TS required equipment operability, but no examples of inadequate
operability determinations were observed.
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i Report Details
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Summarv of Plant Status
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This special inspection was performed while LaSalle Unit 1 was in Mode 4 - Cold !
Shutdown, and LaSalle Unit 2 was in Mode 5 - Refueling.
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1. Ooerations
- 01 Conduct of Operations )
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01.1 General Comments
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The inspectors reviewed two events associated with the licensee's program !
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for taking equipment out of service (OOS). In one event, a non-cited !
violation (NCV) of a Technical Specification (TS) action statement occurred )
as the result of an OOS error. This event also involved the misinterpretation
l of a TS time clock by on-duty operators. The other event, an unintended -
i entry into a TS action statement, was also the result of an OOS error. Both
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OOS errors were caused by inadequate preparation and review by licensed
operators.
- 01.2 Failure to Meet Secondary Containment isolation Reauirements
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l a. Insoection Scone (92901)
On September 28,1996, the licensee identified that irradiated fuel l
l movement was conducted while Containment Purge and Ventilation system ;
valves were in a condition not in conformance with the applicable TS
, requirements. The inspectors performed a special, regional initiative, i
followup inspection using the guidance of Inspection Procedure 92901. The :
inspectors interviewed the involved plant staff, and reviewed the j
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procedures, work instructions, and log entries associated with the event. i
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- b. _Qhlervations and Findinas
At the time of the event, LaSalle Unit 2 was in Mode 5 and the core was
being altered in that irradiated fuel was being unloaded. TS 3.6.5.2 required
that secondary containment ventilation system automatic isolation valves be
operable during core alterations. TS 3.0.5 required that both primary and
backup power supply be available to motor operated valves (MOV) in order
for them to be considered operable in Mode 4 or Mode 5. Isolation valve
2VOO38 in the Containment Purge and Ventilation system was an automatic
isolation MOV which received its backup power from the "0" Emergency
Diesel Generator (EDG).
As implemented by the licensee's work control program, TS 3.6.5.2 required
that isolation valve 2VOO37 be shut and deenergized if isolation valve
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2VOO38 was inoperable during core alterations. This action statement
ensured that at least one containment isolation valve was closed and could .
not be inadvertently opened. This action statement condition was required
to be completed within eight (8) hours of the time that 2VOO38 became
inoperable (8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> " time clock").
The "0" EDG was taken out of service (OOS) at 4:11 a.m. on September 26,
1996 using work package OOS 960007844. The SRO who prepared the
initial OOS package recognized that isolation valve 2VOO38 was rendered
functional but inoperable by removing the "O" EDG from service, and the
OOS specified that 2VQO37 was to be closed and its power supply breaker
opened as required by TS 3.6.5.2. These actions were completed within the
TS required 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time clock.
On September 28,1996, a partial " lifting * of the OOS was initiated so that
the EDG oil system could be warmed up. The EDG was not rendered
operable, so 2VQO37 should have been left closed and deenergized. This
was not recognized by the licensed senior reactor operators (SROs) who
prepared and reviewed the work package for the partiallifting of the OOS.
As a result of their inadequate attention to detail and poor review, the
portion of OOS 960007844 applicable to the power supply breaker for
2VOO37 was cleared, and the breaker was closed.
A reactor operator clearing a separate portion of OOS 960007844 identified
that 2VOO37 should not have been reenergized. The power supply breaker
for 2VOO37 was reopened approximately four (4) hours after it was closed.
Both 2VOO37 and 2VOO38 remained in the closed position during the 4
hours that 2VQO38 was energized.
In performing their initial review of this event, the on-duty operators
characterized the reenergization of 2VQO37 as an inadvertent entry into a
TS action statement rather than as a TS violation. This error was made
because they incorrectly applied the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> time clock to the point in time
when the power supply breaker for 2VOO37 was closed rather than the
point in time when valve 2VOO38 was rendered inoperable. This error led to
a slight delay in reestablishing TS compliance. Plant Management identified
this error, and recharacterized the event correctly.
The licensee implemented the following corrective actions:
All SROs were briefed about the correct imp!smentation of TS action
statement time clocks.
Remedial training was provided for the involved personnel.
A special written examination was given to all station SROs to assess
their performance at the interpretation of T3 requirements. SROs
who did poorly on this test were provided remedial training and were
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required to take a second written examination to demonstrate !
proficiency at interpreting TSs. Any SRO who failed both
examinations was to be administratively relieved from their licensed l
duties.
An ongoing program of refresher training was established to address
the weaknesses identified through review of the completed special
examinations.
An additional SRO review of OOSs was added.
The licensee-identified and -corrected failure to maintain 2VQO37 closed and
deenergized, as required by TS 3.6.5.2, is being treated as a non-cited
violation (NCV 50-373/374-96016-01), consistent with Section Vll.B.1 of
the NRC Enforcement Poliev.
The inspectors considered the Reactor Operator's identification of the
inappropriate closure of the 2VQO37 power supply breaker to be an example
of a good questioning attitude.
While interviewing the personnel involved with the partiallifting of OOS
960007844, the inspectors found that normal practice had been to issue
separate OOS control packages for equipment rendered inoperable by work
on EDGs. This normal practice meant that all physical work directly
associated with the EDGs was controlled within one OOS package, and that
TS required action (s) for associated equipment were controlled by one or
more separate, but administratively linked, OOS packages.
The inspectors concluded that the involved SRO's inadequate review of the
partiallifting of OOS 960007844 was in part attributable to an assumption
that any TS associated equipment operability issues would be covered by
separate OOS packages. The licensee had taken informal action to ensure
that separate OOS packages would be used in the future. The inspectors
concluded that the lack of a single, formally documented, consistently
applied, approach to OOS package content for EDG work was a weakness
which contributed to this event. This concern was discussed with the Plant
Operations Superintendent, who informed the inspectors that the polict
would be formally documented.
The inspectors reviewed the special examination which was given to the
plant's licensed SROs. The inspectors concluded that the examination
provided a fair but challenging test of performance in reading and
interpreting TS requirements.
The inspectors discussed the results of the special examination with on-shift
SROs. The inspectors concluded that the use of a special written
examination had been effective in heightening the SROs' sense of ownership
of the need to improve performance in the interpretation of TS requirements.
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The inspectors found no indication of poor safety consciousness on the part ;
of the Operations Staff while reviewing this event. !
c. Conclusipn
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A Non-cited Violation of a TS action statement occurred as the result of l
inadequate attention to detail and poor review of an COS on the part of I
licensed SROs. The inspectors independently determined that the licensee's i
method of issuing OOSs for TS required equipment affected by work on
EDGs lacked formality. The licensee identified and effectively addressed a
weakness in the SRO staff's application of TS time clocks through the use j
of a special written examination and follow-up training. i
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01.3 Reactor Protection Svstem Trio Function inadvertentiv Rendered Inocerable l
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a. Insoection Scone (92901)
The licensee identified that a portion of the Reactor Protection System (RPS)
had inadvertently been rendered inoperable. The inspectors interviewed the
personnel involved, and reviewed the licensee documentation of the event.
b. Observations and Findinas '
At the time of the event on September 26,1996, LaSalle Unit 2 was
undergoing core alterations. TS 3.3.1 required that a minimum of three (3) ;
intermediate Range Monitors (IRMs) be operable for each of the RPS trip
channels while core alterations were in progress. The Action Statement for
TS 3.3.1 required that a RPS trip channel with fewer than the minimum
number of operable IRMs be placed in the " tripped" condition. Placing both
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RPS trip channels in the tripped condition produced a reactor scram. ;
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On September 25,1996, a reactor scram was inserted as part of an OOS ,
associated with work on the scram air header. This action was taken I
independently of the subsequent events.
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On September 26, two OOSs, 960010110 and 960010113, were issued for
work on the Average Power Range Monitor (APRM) channels E and F. As
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part of these OOSs, four (4) RPS trip channel contacts associated with
APRM E and F trip functions were jumpered out of service. The licensed
SROs who prepared and reviewed OOS 960010110 and 960010113 did not :
recognize that the 4 RPS contacts which were jumpered out of service each '
received a trip input from IRM inoperative and IRM high flux trip functions
(IRM channels E, F, G, and H) as well as from the APRM trip function. As a I
result of the jumpers installed on the RPS trip channel contacts, only two
IRM trip functions were operable per RPS trip channel. The failure to have a
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minimum of 3 operable IRM trip functions operable per RPS trip channel
placed the unit in the Action Statements of TS 3.3.1, which required that
both RPS channels be placed in the tripped condition while core alterations
were underway. Fortuitously, this condition was satisfied because of the
reactor scram which had been inserted for the unrelated scram air header
work.
On September 29,1996, a reactor operator reviewing the work necessary to
reset the reactor scram inserted for the scram air header work made the
initial discovery of the inoperable condition of the IRM trip functions E, F, G,
and H. The operations staff promptly initiated Problem identification Form
(PIF) 96-2857 in accordance with plant procedures.
The inspectors reviewed the PlF, and determined that licensee management
properly characterized the significance of the inadvertent entry into the TS
action statement. Licensee corrective actions included:
A review of existing plant conditions was performed to determine
whether the impact of ongoing work was properly characterized and
controlled. No non-compliances were identified, but 12 additional
caution tags or OOS modifications were made to provide increased
assurance that the plant configuration was being adequately
controlled.
Operations Management met with each licensed SRO and discussed
the details of this event and the need to maintain an appropriate level
of attention to detail when preparing and reviewing OOSs.
The lessons learned from this event were incorporated into the
follow-up training associated with the event described in Section
01.2 of this report.
The inspectors considered the licensed operator's identification of the
inoperable status of the IRM trip functions to be an example of a good
questioning attitude. The inspectors considered the SROs' failure to
adequately prepare and review OOS 960010110 and OOS 960010113 to be
examples of human performance problems.
c. Conclusion
An inadvertent entry was made into a TS Action Statement because of
inadequate preparation and review of an OOS for the RPS. Compliance with
the TS was achieved fortuitously. The poor performance on the part of the
SROs involved with this event is similar in nature to that described in
Section 01.2 of this report, and previously identified in one of the problems
described in Section M1.2 of Inspection Report 50-373/374-96007.
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III. Enaineerina
E1 Conduct of Engineering
E1.1 Potential Generic Concern with Air Operated Valves
a. Insoection Scooe (37550)
The licensee notified the NRC on September 28,1996, that air operated
valves (AOV) in the Primary Containment isolation System (PCIS) were
found to have actuator diaphragm areas less than specified in their written
specifications, and that this condition had the potential to affect the AOVs'
operation under accident conditions. The inspectors performed a review of
the licensee's identification and processing of this issue using the guidance
of Inspection Procedure 37550.
b. Observations and Findinas Soecifically Related to AOVs
The licensee implemented a comprehensive program to perform periodic
preventive maintenance (PM) on AOVs in safety and non-safety related
service. The PM program for WKM Model 70-13 actuators included
disassembly of the housing, replacement of worn parts, and resetting of the
actuator spring preload to original vender specifications. For " spring to
close" AOVs, the spring was preloaded (compressed) such that the valve
closed when instrument air was vented from the actuator diaphragm. This
design repositioned the valve to a " fail safe" configuration on a loss of air.
On March 17,1996, during refueling outage L1R07 scheduled PM on AOVs,
the licensee's maintenance and engineering staffs identified that five (5)
AOVs in the Reactor Core isolation Cooling (RCIC) system and two (2) AOVs
in the Primary Containment isolation System (PCIS) exhibited spring preload
characteristics which were not consistent with the vender's written
specifications. All seven AOVs had WKM Model 70-13, size 70 actuators.
The AOVs in the RCIC system were associated with condensate drains. The
PCIS AOVs were two (2) inch diameter, Drywell Floor Drain isolation valves.
The Unit 2 RCIC and PCIS systems contained similar AOVs. Unit 2 was at
power at the time of this discovery.
Licensee procedure LAP-1500-8A, " Initiating A Problem Identification Form"
(PIF), Revision 0, directed that a material which failed to meet drawing or
written specifications, or a material with faulty manufacturing, be
documented on a PIF. Site Engineering immediately recognized that the
actuator springs, or some other material aspect of the actuators, were in
non-conformance with written specifications. A corrective action process
was started, but no PIF was initiated to document this condition which
affected quality. Because of the failure to initiate a PIF, a formal operability
evaluation was not perfocmed for the Unit 2 valves, or for the Unit 1 valves
prior to unit restart. The inspectors considered the failure to initiate a PlF as
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required by plant procedure to be an example of a violation (VIO 50-
373/374-96016-02a) of 10 CFR 50, Appendix B, Criterion V.
Site Engineering diligently pursued the AOV problem during the last two
weeks of March 1996 by directing that additional tests of WKM Model 70-
13 actuators be performed. Based upon these additional tests, Site
i Engineering concluded that WKM Model 70-13 actuators in sizes 35,70,
140, and 280 all exhibited spring closure forces less than expected when
the spring preload was set using the vender's written specifications. This
condition was the result of the actual effective diaphragm area (EDA) being
less than that specified in the written specifications for the valves. The EDA
affected the spring force because the spring preload was set by applying a
known air pressure against the assumed area of the diaphragm and adjusting
the spring compression for desired valve stem travel. For the known air
pressure, a smaller than anticipated spring pPJoad compression was
achieved because of the smaller than assumed EDA. The spring preload was
significant because it was relied upon to overcome dynamic loads on the
valve's disc and friction loads on the valve's stem if the valve was called
upon to close under accident conditions.
Site engineering determined that the discrepancy in EDAs was caused by:
manufacturing / design error in the diaphragm dimensi?ns,
installation orientation of the diaphragm and actuator housing which
caused hard contact between the diaphragm components and the
housing,
a tendency for the diaphragm to stretch when air was applied to one
side.
Site Engineering determined that the Model 70-13 actuator was initially sold
by Black, Syvals and Bryson (BSB). BSB was purchased by WKM. WKM
supplied most of the Model 70-13 actuator fitted AOVs at LaSalle. WKM
was purchased by Muesco. Muesco supplied LaSalle with two (2) Model
70-13 actuator AOVs. Finally, The Anchor Darling Valve Company (ADVC)
purchased Muesco and became the vender of record.
On April 1,1996, Site Engineering issued Nuclear Design Information
Transmittal (NDIT) LS-0252, which provided corrective actions to be taken
in repairing the Unit 1 and Unit 2 RCIC AOVs which had WKM Model 70-13
actuators in order to bring them into conformance with their written
specifications. On April 2,1996, Site Engineering issued NDIT LS-0253,
which provided corrective actions to be taken in repairing the Unit 1 and Unit
2 Drywell Floor Drain PCIS valves to bring them into conformance with their
written specifications. These NDITs included screening performed in
accordance with 10 CFR 50.59, which determined that the repairs would
improve performance of the valves' safety function.
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Licensee procedure LAP-220-5, " Equipment Operability Determination," ,
Revision 3, required that any operability issues, and supporting
documentation, were to be provided to the Shift Engineer (SE). The
procedure further required that the SE was to become familiar with the
operability issue, was to ensure that a PlF was generated, and was to
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perform and document an operability determination on the PlF. Site
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Engineering initiated instructions to repair the valve actuators, but the plant *
- did not initiate an operability assessment to determine whether these
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components could perform their safety function under accident conditions '
prior to repair. The inspectors considered the failure to perform an }
operability determination as required by plant procedure to be an example of i
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a violation (VIO 50-373/374-96016-02b) of 10 CFR 50, Appendix B, [
- Criterion V,
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The licensee contacted ADVC to obtain assistance in evaluating the as-found ,
conditions and to obtain guidance on corrective actions as early as March l
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23,1996. Additional contact between the licensee and ADVC took place as i
each party continued to evaluate the as-found EDAs. l
k On April 10,1996, Site Engineering initiated work scope revision forms to
i repair all 26 (total) PCIS AOVs fitted with Model 70-13 actuators during the
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current or upcoming outages. These valves wern in the Recirculation .
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l Sample Line System, the Drywell Floor Drain System, the Drywell Equipment
- Drain System, and the Instrument Nitrogen System. The affected valves
j ranged in size from 3/4" to 2". l
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On May 7,1996, ADVC committed to the licensee that they would review !
the as-found condition for reportability under 10 CFR 21. In accordance !
with licensee procedures, LaSalle initiated a Nuclear Tracking System (NTS) !
- item to monitor the status of ADVC's 10 CFR 21 review.
On May 7 and 8,1996, the LaSalle AOV component engineer presented the
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WKM 70-13 AOV actuator preliminary findings to a Institute of Nuclear ,
Power Operations (INPO) AOV working group. l
On September 20,1996, LaSalle and ADVC completed enough evaluatory !
studies to proceed with corrective action plans for the effected PCIS valves. ,
i On September 26,1996, Site Engineering determined that NDITs were not i
< appropriate for the scope of corrective modifications required for the AOVs l
with Model 70-13 actuators. The licensee commenced preparation of i
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Design Change Packages (DCPs) to implement the required changes.
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On September 28,1996, preliminary calculations performed by Site
Engineering indicated that some of the affected PCIS AOVs might not close ;
under all accident conditions if the most conservative assumptions were
made about how the original spring preloads were set. A PIF was promptly !
generated and the AOVs with WKM Model 70-13 actuators were declared j
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inoperable. The NRC was notified in accordance with 10 CFR 50.72(b)(2).
On October 4,1996, the licensee made an interim notification regarding the
potential generic concern with the WKM Model 70-13 AOV actuators in
accordance with 10 CFR 21 Sections 21.1(b), 21.3a(3), and 21.3d(4). i
At the conclusion of the special inspection, the licensee was developing ,
4 plans to test the as-found spring preload for those PCIS AOVs which were !
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not repaired during L1R07. This testing was necessary to determine
i whether the springs had been originally set with more preload compression
- than the licensee achieved using the vender's procedure. The licensee
informed the inspectors that individual 10 CFR 50.72 notifications would be
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issued for any valves which were found to have spring preloads which were
inadequate to ensure performance of the AOV's safety function. The
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inspectors considered the regulatory significance of the PCIS AOVs' as- l
found spring preloads to be an unresolved item (URI 50-373/374-96016-03)
pending receipt of the licensee's test results.
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l The inspectors concluded that the licensee's AOV PM initiative was
thorough and well conceived. The inspectors further concluded that Site
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Engineering diligently pursued resolution of the technical aspects of the '
- problem with WKM Model 70-13 AOV actuators. The inspectors identified
, that Site Engineering and all other plant staff involved in the identification,
evaluation, processing, and scheduling of the AOV issue failed to adequately
implement the plant's procedures for identifying and assessing non-
conformances. The inspectors found no indication that the failure to initiate
a PIF or perform an operability determination was the result of deliberate
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c. Additional Observations and Findinas Related to AOVs
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j i. Weakness in the Bases for Determinino Comoonent Ooerability
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The inspectors interviewed various members of the plant staff to determine -
why the difficulties with obtaining AOV actuator conformance to written i
specifications were not documented on a PlF or identified as an operability
issue. On three occasions, plant staff indicated that they believed the AOVs
- were operable because they had passed TS required surveillances for local
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leak rates and for stroke times. The inspectors reviewed plant procedure
LAP-220-5 " Equipment Operability Determination," Revision 3, Paragraph
j 2.m.(1)(l), and found that the procedure appeared to stress conformance to
, TS specified functions as being the principal indication of operability.
The inspectors concluded that some plant staff were interpreting LAP-220-5
to mean that a SSC was presumed operable if it passed its TS surveillances. i
Thic interpretation was inconsistent with the NRC's documented position '
that TS surveillances provide indication of operability, but do not assure
conformance to all design requirements. The NRC's guidance was ,
documented in Generic Letter (GL) 91-18 " Guidance on Non-Conforming
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Conditions and Operability." An example of the need to look beyond TS
surveillances when making operability determinations was provided in GL 89-
10 " Motor Operated Valves."
The inspectors concluded that the views expressed by plant staff members
and the lack of clarity in LAP-220-5 were indicative of a weakness in the
licensee's program for identifying and assessing non-conformances.
ii. Review of Comoteted and Ooen Operability Determinations
The inspectors reviewed four PlFs which described conditions for which Site
Engineering was providing technical assistance in completing an on-going i
operability determination. The inspectors selected six PIFs which
documented conditions for which completed operability determinations had
been made. No examples of poor safety consciousness or inadequate
technical bases for operability determinations were identified.
iii. Review of NDITs
The inspectors noted that Site Engineering's initial response to the problems
identified with WKM AOV actuators was to provide corrective actions via '
NDITs. Two NDITs were prepared and issued without performing a rigorous
assessment of the impact of the as-found condition on valve operability.
The inspectors reviewed Commonwealth Edison Company procedure NEP-
12-03, " Nuclear Design information Transmittals (NDIT)," Revision O. NEP-
12-03 clearly indicated that NDITs were to be used to transmit data in a
controlled manner and were not to be used as design or installation
documents. The inspectors reviewed NEP-12-03LA, "LaSalle Nuclear Design
Information Transmittals (NDIT) Site Appendix," Revision 2. This Appendix
expanded the scope of NDIT usage at LaSalle to include minor changes to
safety-related design documents.
The inspectors reviewed the index of LaSalle NDITs and selected 10 that
dealt with safety-related or important to safety equipment. The inspectors
performed a cursory review to determine if the 10 NDITs dealt with
conditions for which a PlF should have been issued and an operability l
assessment would have been required. No examples were found. The i
inspectors reviewed the index of NDITs to see whether any of the
descriptions appeared to involve " intent" changes to safety-related design
documents for which TS 6.2 would have required additional review and
approval by the on-site review committee. No examples were found within
the time frame available during this special inspection.
The inspectors were concerned with Appendix NEP-12-03LA changing the
intent of the NDIT process defined in corporate procedure NEP-12-03. This
matter is condidered to be an inspection followup item (IFl 50-373/374-
96016-04) pending further NRC review.
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d. Conclusiong
The inspectors concluded that the licensee was proactive and thorough in
identifying and addressing the technical aspects of a concern with the EDA
of WKM Model 70-13 AOV actuators. The inspectors identified a violation
of 10 CFR 50, Appendix B, Criterion V, in that the licensee did not promptly
, identify the AOV concern on a PlF or perform an operability assessment as
required by the plant's procedures. The inspectors reviewed the licensee's
performance of other operability determinations. No additional violations
were identified, but a potential weakness in implementing the NRC's generic
guidance for determining the operability of TS required equipment was
found. An unresolved item was opened to track the results of licensee "as-
,
)
found" testing of AOV actuator spring preloads. ,
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X1 Exit Meeting Summary
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The inspectors presented the results of the inspection to members of licensee management
at the conclusion of the inspection on October 8,1996. The licensee acknowledged the l
findings presented.
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The inspectors asked the licensee if any materials examined during the inspection should 1
l be considered proprietary. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED l
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Licensee:
'W. Subalusky, Site Vice President
"D. Ray, Station Manager
- L. Guthrie, Operations Manager
"R. Fairbank, System Engineering Supervisor
"J. Burns, Regulatory Assurance Supervisor
- At exit meeting on October 8,1996.
INSPECTION PROCEDURES USED
IP 37550 Engineering
IP 92901 Followup - Operations
ITEMS OPENED, CLOSED, AND DISCUSSED
Ooened
373/374-96016-01 NCV Secondary Containment TS Action Statement not met ,
for Ventilation Valve. {
373/374-96016-02a VIO Failure to promptly initiate a PlF for AOVs. !
373/374-96016-02b VIO Failure to promptly perform an Operability assessment )
of AOVs. l
373/374-96016-03 URI PCIS AOVs with undersized actuators.
373/374-96016-04 IFl Scope change in NDIT procedure.
LIST OF ACRONYMS USED
ADVC Anchor Darling Valve Company
AOV Air Operated Valve
APRM Average Power Range Monitor j
BSB Black, Syvals and Bryson
EDG Emergency Diesel Generator i
GL Generic Letter !
INPO Institute of Nuclear Power Operations ;
IRM Intermeolate Range Monitors l
NCV Non-cited Valation
NDIT Nuclear Dehn Information Transmittal
NRC Nuclear R9gulatory Commission
NTS Nuclear Tracking System '
OOS Out of Service
PCIS Primary Containment isolation System
PlF Problem identification Form
PM Preventive Maintenance
RCIC Reactor Core isolation Cooling
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SE Shift Engineer
SRO Senior Reactor Operator
TS Technical Specification
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