ML20141K478
| ML20141K478 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 05/16/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20141K472 | List: |
| References | |
| 50-461-97-06, 50-461-97-6, NUDOCS 9705290167 | |
| Download: ML20141K478 (60) | |
See also: IR 05000461/1997006
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U.S. NUCLEAR REGULATORY COMMISSION
REGION lil
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Docket Nos:
50-461
License Nos:
Report No:
50-461/97006 (DRP)
Licensee:
Illinois Power Company
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Facility:
Clinton Power Station
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Location:
Route 54 West
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Clinton, IL 61727
Dates:
February 15 - March 29,1997
Inspectors:
F.D. Brown, Acting Senior Resident inspector
K.K. Stoedter, Resident inspector
R.A. Langstaff, Resident inspector
D.E. Zemel, Resident inspector - IDNS
Approved by:
Geoffrey C. Wright, Chief
Clinton Oversight Team
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9705290167 970516
DR
ADOCK 05000461
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EXECUTIVE SUMMARY
Clinton Power Station
NRC Inspection Report 50-461/97006 (DRP)
This integrated inspection included aspects of licensee operations, engineering,
maintenance, and plant support. The report covers a 7-week period of resident inspection.
Operations
The licensee implemented a revised procedure adherence policy instruction and
performed extensive training on the new requirements. The inspectors did not
identify any compliance problems with the new policy, but were concerned that the
applicability of Prerequisites for independent Sections of procedures was to be
determined by procedure users rather than procedure preparers and reviewers as
recommended in ANSI N18.7. The inspectors considered the consistency and
clarity of the new policy instructions to be an improvement, but noted one
weakness; a non-documented "noted in" concept was being implemented by plant
staff shortly after the new policy was implemented. (Section 01.2)
Maintenance
The prejob brief, engine preparation, and EDG quick start surveillance were
accomplished in a professional manner. (Section M1.2)
The Division I emergency core cooling system surveillance was performed in a
thorough and careful manner. The licensee's expectations for the performance of
independent verification were adhered to during the surveillance, and procedural
adherence was noted. Although the procedure required multiple entries into high
radiation and high contamination areas, no radworker deficiencies were identified.
(Section M1.3)
The surveillance procedure for RHR Containment Pressure Instrument Calibration
was properly implemented in a professional manner. (Section M1.4)
The inspectors observed performance of a surveillance test on the Drywell Purge
portion of the Containment Ventilation System. One procedural violation was
observed. (Section M1.5)
The inspectors reviewed MWR D60080, the work authorization and coordination
document for the modification of the outboard check valve in the "A" Feedwater
line, and identified that it contained very poor documentation of the completion of
this safety-related activity. Seven examples of a violation of the Technical
Specification for Procedures were identified. (Section M1.6)
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Landing of the Drywell head was performed in an appropriate manner. No technical
or radiological control noncompliances were identified, but poor communications
resulted in an avoidable entry into a high contamination area. (Section M1.7)
The inspectors determined that the RCIC System was being maintained per the
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vendors preventive maintenance recommendations. Lubrication of the pump and
turbine was also being maintained in accordance with vendor recommendations.
No significant problems with material condition or preventive maintenance were
identified. (Section M1.8)
The licensee performed an effective follow-up on an LER from another utility, and
identified that a RR system drain line had been damaged by installation of two
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freeze seals, in close proximity to each other, during the current outage. The
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inspectors reviewed the installation of these freeze seals and identified that the
procedure, training, and supervisory control of this evolution was inadequate to
prevent damage to the reactor coolant system boundary. One violation of NRC
requirements was identified. (Section M3.1)
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The inspectors identified that the licensee procedure for performing reactor coolant
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system leakage tests was inadequate in that it allowed test pressures significantly
below those experienced during normal power operations. The procedure had been
used during the 1995 and the current refueling outages. One violation of NRC
requirements was identified. (Section M3.2)
Enaineerina
The licensee's engineering organization identified that some TS required quantitative
values had been incorporated directly into surveillance procedures without
consideration of instrument inaccuracies. The inspectors considered the original
practice of using uncorrected TS values in surveillance procedures to have been
poor, but concluded that the identification and aggressive response to this issue
were an example of a better questioning attitude and safety focus within
Engineering. An unresolved item was opened pending inspector review of the
licensee's evaluations of procedural adequacy and past operability. (Section E1.1)
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Report Details
Summarv of Plant Status
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The plant entered operational Mode 4 at 4:00 pm on February 15,1997. The licensee
continued to perform outage activities and was performing startup preparation activities at
the conclusion of the inspection period.
1. Ooerations
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Conduct of Operations
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01.1 General Comments
The inspectors observed the conduct of Operations staff performing normal outage
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related control room and surveillance activities.
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Specific observetions for surveillance activities are contained in Section M1 of this
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report. Activities associated with surveillance procedure CPS 9061.04,
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" Containment /Drywell isolation Auto Actuation," were of concern to the inspectors.
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A specialinspection was conducted to review these activities. The results of the
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special inspection will be documented in Inspection Report (IR) 50-461/97007.
01.2 Procedure Adherence Policy and Proaram
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a.
Insoection Scone
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The inspectors followed the licensee's actions to correct procedural adherence and
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adequacy problems.
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b.
Observations and Findinas
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The adequacy of the licensee's procedure adherence policy and programs had been
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identified as a concern by the NRC in IR 50-461/96010,50-461/96011, and 50-
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461/96015. The licensee implemented revised procedure adherence policy and
program instructions, CPS 1005.15, " Procedure Use and Adherence," near the end
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of this reporting period. The licensee informed the inspectors that this new policy
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and the implementing program requirements were being fully integrated into all
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upper tier procedures such as the procedures for Conduct of Operations and for
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Conduct of Maintenance. The inspectors considered this to be an improvement in
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program consistency and clarity.
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The inspectors reviewed CPS 1005.15, and identified one concern. American
National Standard Institute (ANSI) N18.7, " Administrative Controls and Quality
Assurance for the Operational Phase of Nuclear Power Plants," a standard to which
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the licensee was committed, Paragraph 5.3(4), stated " Prerequisites applicable only
to certain sections of a procedure should be so identified." The inspectors noted
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that CPS 1005.15 stated that when independent performance of procedure
sections was performed, it was the responsibility of the performer to determine
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which prerequisites were applicable to that section. The inspectors were concerned
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that the licensee's approach provided less positive assurance that appropriate
prerequisites would be identified than that recommended by ANSI N18.7. The
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licensee acknowledged the inspectors' concom, and stated that the issue would be
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evaluated during a revision to 1005.15 scheduled for after unit start-up.
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The licensee performed extensive training of all plant staff in the new policy and the
new program instructions prior to implementing the changes. The inspectors
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observed some of these training sessions and had no negative observations,
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The inspectors identified one concern while observing the Shift Supervisor (SS)
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turn over on the morning of March 31,1997. The inspectors noted the off-going
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SS mention that information had been "noted in" to CPS 9000.010002, " Control
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Room Operator Surveillance Log - Mode 4,5 Data Sheet," because plant conditions
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were different than expected. The inspectors followed-up this observation by
discussing the "noted in" concept with operators and Operations management, and
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by reviewing the copy of CPS 9000.01DOO2 for March 31,1997. The inspectors
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were told that "noted in" was understood to be the process of adding extra
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information to a procedure. The inspectors were told that this concept was not
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based on written guidance, but had been developed from comments made during
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the procedure adherence training and during verba! conversations between SSs and
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plant management. The inspectors determined that CPS 9000.010002 was not a
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procedure required by NRC regulations or the facility license, so the "noted in" sign-
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offs were not a regulatory issue in this case. The inspectors expressed to licensee
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management the concem that verbally communicated procedure adherence
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guidance was subject to misunderstanding and misapplication. Licensee
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management acknowledged the inspectors' concern. The inspectors did not
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identify any examples of procedural violations associated with "noted in" changes
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to safety-related procedures between March 31,1997 and the end of this
inspection period.
In addition to the new procedure adherence policy, the licensee also implemented a
new program for performing and controlling temporary procedure changes. The
licensee informed the inspectors that the intent of this new program was to
facilitate efficient, regulatory compliant, procedure changes when field conditions
differed from those for which a procedure was written.
c. Conclusions
The licensee implemented a revised procedure adherence policy instruction and
performed extensive training on the new requirements. The inspectors did not
identify any compliance problems with the new policy, but were concerned that the
applicability of prerequisites for independent sections of procedures was to be
determined by procedure users rather than procedure preparers and reviewers as
recommended in ANSI N18.7. The inspectors considered the consistency and
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clarity of the new policy instructions to be an improvement, but noted one
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weakness in that a non-documented "noted in" concept was being implemented by
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plant staff shortly after the new policy was implemented.
II. Maintenance
M1
Conduct of Maintenance
M1.1 General Comments
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a.
insoection Scone (61726)
Portions of the following surveillance activities were observed or reviewed by the
inspectors.
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-CPS 9030.01D028
ATM Channel Functional and Calibration Check for RHR
Containment Pressure Instruments E12-N662A, B, C,
and D
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-CPS 9052.01
LPCS/RHR A Pumps and LPCS/RHR A Water Leg Pump
Operability
-CPS 9053.05
RHR/LPCS Valve Operability (Shutdown)
-CPS 9059.01
Reactor Coolant System Leakage Test
-CPS 9061.03
Containment /Drywell isolation Valve 3 Month
Operability
-CPS 9061.04
Containment /Drywell isolation Auto Actuation
-CPS 9080.01
Emergency Diesel Generator 1 A/1B Operability (Both
the Division I and ll EDG tests were observed)
-CPS 9432.08
Main Steam Line ambient temperature E31-N604C/D
Channel Calibration
b.
Observations and Findinas
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The inspectors found the work performed under these activities to be generally
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acceptable, with procedures present and in use. Comments for specific work
activities are discussed in further detail below. The performance of CPS 9061.04 is
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discussed in a separate inspection report (50-461/97007) due to the significant
issues that were identified during its performance.
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M1.2 Division i Diesel Generator SurveillanGR
a.
Insoection Scone (61726)
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The inspectors observed the pre-brief, engine preparations, and accomplishment of
the surveillance for quick starting the Division 1 Diesel Generator (DG) (CPS
9080.01).
b.
Observations and Findinos
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A thorough pre-job brief was performed, including contingency planning for
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potential equipment problems.
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The preparations for starting the DG were accomplished by three operators and
were in accordance with the procedure. Particular attention was placed on barring
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over the DG no more that two revolutions. Recent concerns with barring the
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engine and its preconditioning effect were well recognized by the operators.
The operators took 25 minutes to accomplish the procedural steps when the DG
was in lockout and unavailable. The procedure contained a caution statement
directing the time in lockout be minimized. In this case, the operators were
knowledgeable of the required steps and performed the steps in a prompt but
controlled manner.
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The surveillance procedure was written for quick starting the DG and loading it to
the electrical grid. In this case (plant was in Mode 4), the loading was not required.
The procedure contained adequate direction to mark specific DG loading steps as
"not applicable." In addition, the procedure addressed other steps that may not be
required and how to annotate those circumstances. The surveillance was
accomplished without problem and the DG was restored to operable status.
c.
Conclusions
The pre-brief, engine preparation, and EDG quick start surveillance were
accomplished in a professional manner.
M1.3 Performance of Low Pressure Core Sorav and Residual Heat Removal A
Surveillances
a.
Insoection Scone (61726)
The inspectors observed the pre-brief and performance of CPS 9052.01 and CPS
9053.05.
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b.
Observations and Findinas
Briefings on both of these surveillance procedures were detailed and discussed the
coordination needed for proper surveillance performance and possible
contingencies. Technical specification actions were also discussed to ensure that
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actions were entered at the appropriate time.
Although the surveillance procedure was complex and required that activities be
performed by several groups, coordination was generally good. For example, two
operators were allocated to this activity since several of the operations had to be
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performed in high radiation /high contamination areas. The use of two people
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ensured that additional time was not spent having one operator traverse in and out
of these areas in order to perform the necessary procedure steps.
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The inspector noted good use of the procedure by both operators. The licensee's
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recently implemented self checking techniques were employed as a method of
verifying the proper equipment had been identified prior to performing any
manipulations. In addition, the second operator allocated to this activity served as
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a second checker which further confirmed that the proper equipment was operated
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and ensured that the self checking methods were properly performed.
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c.
Conclusions
The Division i emergency core cooling system surveillance was performed in a
thorough and careful manner. The licensee's expectations for the performance of
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independent verification were adhered to during the surveillance, and procedural
adherence was noted. Although the procedure required multiple entries into high
radiation and high contamination areas, no radworker deficiencies were identified.
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M1.4 RHR Containment Pressure Instrument Calibration
a.
Insoection Scone (61726)
The insp?ctors observed parformance of portions of surveillance procedure CPS
9030.01DO28 "ATM Channel Functional and Calibration Check for RHR
Containment Pressure Instruments E12-N662A, B, C, and D."
b.
Observations and Findinas
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The inspectors found the Controls and Instrumentation (C&l) technician to be very
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familiar with the procedure being implemented. The procedure was appropriately
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released for work by plant operations. The C&l technician completed the procedure
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as written, and was familiar with the layout and location of the panels and alarms
associated with the test.
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c.
Conclusions
The surveillance procedure for RHR Containment Pressure Instrument Calibration
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was properly implemented in a professional manner.
M1.5 Containment /Drvwell Ventilation System Surveillance
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a.
Insection Scone
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The inspectors observed performance of the Drywell Purge System portion of
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surveillance test CPS 2104.02, "VQ/RA Charcoal Absorber Leak Test." The
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Drywell Purge System was a part of the Containment Ventilation System as
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described in CPS 3408.01, " Containment Building /Drywell HVAC VR/VO." As the
result of this operation, the inspectors initiated a review of the test methodology for
this surveillance.
b.
Observations and Findinas
During observation of CPS 2104.02 on February 27,1997, the inspectors
witnessed licensee staff make an error while testing filter train OVQ07FB. The CPS
2104.02 test used Halide to verify the effectiveness of systems' absorber banks.
The CPS 2104.02 test was performed in conjunction with test procedure CPS No.
2104.01, " Ventilation Filter Train Testing". Both procedures required the use of
two sample tubes which were installed in sample ports in the filter train housing.
The licensee staff placed all four sample lines in the sample ports prior to
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connecting the other end of the sample lines to the test equipment used for the two
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types of test. The test personnel then connected the correct up stream (U/S)
sample line to the U/S sample connection on the test equipment. They then
connected a down stream (D/S) sample line to the D/S sample connection on the
test equipment. The licensee staff then attempted to perform the CPS 2104.02
test, but the results were unreasonable. The test personnel concluded that the
sample lines were not properly connected to the test equipment, in that the sample
line connected to the D/S port of the test equipment was installed in the wrong
location in the system (this sample line had been staged for the other surveillance
procedure). The test personnel corrected the sample line connections and
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successfully completed the test.
The inspectors noted that the original D/S sample line was located immediately
after the absorber bank rather than further down stream as required. This error
was potentially significant because Halide which bypassed the absorber had not had
a chance to be uniformly distributed in the airstream. This condition could have
resulted in conservative or non-conservative test results. Fortuitously, the test
personnel recognized the results were unreasonable. This led them to identify their
error.
The test personnel stopped performance of tests using CPS 2104.02 and originated
CR No.1-97-02-276. This CR identified that the procedure, as written, was
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inadequate to successfully perform the surveillance test in that it did not specify
that the D/S sample line be connected to the D/S sample connection on the Halide
test equipment. The inspectors considered the use of an inadequate test procedure
to perform a surveillance test on a portion of the Containment Ventilation System
to have been a violation (50-461/97006-01) of T.S. 5.4.1.
In following-up the surveillance observations, the inspectors noted that an unusually
large range of system flows was specified in the surveillance acceptance criteria.
The inspectors initiated a review of the USAR and original system test
documentation to assess the validity of the surveillance. This effort had not
completed at the end of the inspection period and will be tracked as an inspector
Follow-up item (50 461/97006-02).
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c.
Conclusions
The inspectors observed performance of a surveillance test on the Drywell Purge
portion of the Containment Ventilation System. One procedural violation was
observed.
M1.6 Performance of Feedwater Check Valve Modifications
a.
Insoection Scone (62703)
The inspectors reviewed the Maintenance Work Request (MWR) package for the
modification of the outboard check valve in the "A" Feedwater line following
completion of allidentified maintenance activities, but prior to close-out of the
MWR.
o.
Observations and Findinas
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The inspectors obtained a partial copy of MWR D60080, "RF6 Rework Valve
1B21F032A Due to LLRT Failure," on or about March 19,1997 to review the
status of three discrepancies discussed by the licensee at a morning status
meeting. Final reassembly and post modification testing of 1821F032A had been
completed on or about March 8,1997. The inspectors performed a cursory review
of this partial copy of the MWR and identified that at least 40 of the approximately
110 job steps (J/Ss) were not signed as being completed. The inspectors
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concluded that no more than four of these J/S should have been unsigned based
upon the status of the valve work as described in the licensee's status meetings.
On March 31,1997 the inspectors performed a more thorough review of the
MWR D60080 work package. During this review, the inspectors identified that 22
J/Ss had still not been signed for. The other 18 J/S had been signed as complete
or marked "N/A" (non-applicable) between March 19 and March 31,1997. Among
the J/Ss still unsigned, were those for proper check valve reassembly.
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The inspectors reviewed the licensee's upper tier procedures CPS 1501.02,
" Conduct of Maintenance," and CPS 1029.01, " Preparation and Routing of
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Maintenance Work Documents" to determine whether the licensee's programs were
in compliance with the NRC requirements and the utility's licensing commitments.
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The inspectors identified that CPS 1501.02 and CPS 1029.01 did not require
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MWRs to be prepared or performed to the standards specified for procedures. The
inspectors also noted that MWRs were not required to be performed as written, and
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therefore would not satisfy the requirements for documented instructions or
drawings required for maintenance activities that could affect the performance of
safety-related equipment.
The inspectors discussed the above observations with managers of the
Maintenance Department, who informed the inspectors that the licensee did not
consider MRWs to be documented instructions, procedures or drawings as
described in the NRC requirements. The inspectors noted, and plant staff
concurred, that this distinction was not clearly delineated in CPS 1501.02 and CPS
1029.01. The inspectors also noted, and plant staff again concurred, that the use
of MWRs as a work authorization and coordination tool did not relieve the licensee
from the requirement of controlling maintenance activities which could affect the
performance of safety-related equipment by use of written procedures, documented
instructions, or drawings. The plant staff assured the inspectors that the
maintenance planners who prepared MWRs were aware of the intended function of
MWRs and the need to reference approved written procedures, documented
instructions, or drawings for each MWR J/S.
Comoliance issues
The inspectors reviewed MWR D60080 and the rest of the available work package
for the 1821F032A modification to assess compliance with NRC requirements and
the utility's licensing commitments, and identified a lack of attention to detail in the
licensee's handling of the MWR.
The inspectors identified that J/Ss were not signed-off as work was completed, as
required by CPS 1501.02, step 8.1.4.8. MWR D60080, J/Ss 77 through 87
covered the assembly of 1821F032A. These job steps were not signed for in the
MWR. J/S 93 required that maintenance mechanics reassemble 1B21F032A in
accordan::e with CPS 8120.04 and the vendor manual. J/S 93 had not been signed
by a maintenance mechanic. Failure to sign for J/Ss 77 through 87 and 93 as work
was performed was considered to be an example of a violation (50-461/97006-
03a) of T.S. 5.4.1.
The inspectors identified that procedurally required verifications had not been made.
MWR D60080, J/S 93 required that maintenance mechanics reassemble
1B21F032A in accordance with CPS 8120.04 and the vendor manual. A copy of
CPS 8120.04C001, the checklist for verification signatures associated with CPS
8120.04, had been prepared for use as part of MWR D60080 on February 10,
1997. CPS 8120.04, Steps 8.9.8, 8.11.1.12, 8.11.2.15, 8.12.5, 8.12.7, and
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8.15.11 were neither marked as non-applicable nor initialed and dated on the
working copy of CPS 8120.04 COO 1, as required by CPS 1005.01, " Procedures and
Documents," Step 8.1.1. Failure to perform procedural steps as directed was
considered to be an example of a violation (50461/97006-03b) of T.S. 5.4.1.
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MWR D60080, J/S 42 required that 1821F032A be inspected in accordance with
the applicable sections of CPS 8120.34, " Check Valve inspection," with
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notifications and measurement data to be documented on checklist 8120.34 COO 1.
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The hspectors confirmed that J/S 42 had been initialed as complete, but found that
the cr.py of 8120.34 COO 1 associated with this J/S was not completely filled in and
was not signed by the individual who performed the work. Failure to adequately
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complete and sign for the performance of work, as required by CPS 8120.34, was
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considered to be an example of a violation (50461/97006-03c) of T.S. 5.4.1.
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The inspectors identified that a J/S had been inappropriately signed for. MWR
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D60080, Job Step 46 required that Operations be notified after pipe hangers on line
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11SO1 A were cut. A late entry sign-off of this J/S was completed on 3/26/97.
CPS 1501.02, Step 2.1.3 allows such late entries as long as they are based on
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objective evidence of the performance of the J/S. The inspectors requested a copy
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of the objective evidence supporting the late entry sign-off of J/S 46. The licensee
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informed the inspectors that there was no supporting evidence and that the step
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should have been marked Non-Applicable because the hangers were not removed
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from line 11SO1 A. The licensee cited the reason for this error as being inattention
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to detail. The inspectors considered the late entry sign-off of J/S 46, without
supporting objective evidence of the completion of the step to be an example of a
violation (50461/97006-03d) of T.S. 5.4.1.
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The inspectors identified that required rework was not documented and that final
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component rneasurements were not obtained. MWR D60080, J/S 54 required the
fabrication of new actuator shafts for the valve, and specified that the fabricated
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components' dimensions be verified by Quality Verification (QV). The inspectors
reviewed the work package and identified that the record of work activities on the
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MWR continuation sheets documented re-machining of the shafts after the final OV
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verification of component dimensions was performed. The inspectors identified this
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issue to licensee maintenance management. The licensee acknowledged that final
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measurements had not been documented in the work package as specified in the
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MWR. The inspectors concluded that no procedural requirement for obtaining the
final component measurements existed. CPS 1015.02, Step 8.1.4.8, required that
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MWR J/Ss which were reperformed be documented on a copy of CPS
1029.01FOO3. The shaft re-machining was not documented on a copy of CPS
1029.01FOO3. Failure to reperform and document the additional fabrication work
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which changed the actuator shaft dimensions covered by MWR D60080 J/S 54
was considered to be an example of a violation (50-461/97006-03e) of T.S. 5.4.1.
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The inspectors identified that required documentation was not maintained. The
licensee initiated Condition Report (CR) 1-97-03-111 to document the need to
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disposition three deviations from the fabrication drawings for the new internal
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components of 1821F032A. This CR was initiated after the valve had been
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reassembled. CPS 1501.02, " Conduct of Maintenance," Section 8.9, allows " Risk
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Basis" deviations from design if the applicable engineer is contacted at the time of
identification of the deviation and if the engineer provides verbal approval to
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continue with the work. Step 8.9.1.2 of CPS 1501.02 requires that the effort to
contact the engineer and the engineer's approval to continue work shall be
i
documented on CPS 1029.01F010, "MWR Discrepancy List." The inspectors
identified that the applicable copy of CPS 1029.01F010 did not document either an
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effort to contact an engineer regarding the fabric ~ation errors, nor that approval to
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continue with valve reassembly had been obtained. Failure to document
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engineering approval to proceed with the " Risk Basis" reassembly of 1821F032A
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was an example of a violation (50 461/97006-03f) of T.S. 5.4.1. The licensee
informed the inspectors that the required verbal approval had been obtained.
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The inspectors identified that the licensee did nc,t have a process for ensuring that
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components with design deviations were not returned to service prior to resolution
of the deviation. CPS 1501.02, Step 8.9.1.5, requires that " Risk Basis"
!
discrepancy list deviations be resolved prior to declaring the effected equipment
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operable. CPS 1501.02, Step 2.2.19 and CPS 1029.01, Step 4.1 provide similar
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direction, but add that effected equipment should not be returned to service as well
{
as not being declared operable. The inspectors reviewed the three " Risk Basis"
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deviations in the 1821F032A valve work package which had been documented on
CR 1-97-3-111. The deviations were found to have been resolved by Engineering
!
Change Notice (ECN) 30099 on March 14,1997. The inspectors reviewed the
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Daily Plant Outage Reports and concluded that Residual Heat Removal (RHR) "A"
had been considered as operational with flow via the Feedwater "A" line and
1821F032A prior to ECN 30099 being issued. The licensee was requested to
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provide any documentation which contradicted this conclusion, but had not done so
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at the time this report was issued. Returning 1821F032A to service as an
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operable flow path for RHR prior to resolving " Risk Basis" deviations was an
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example of a violation (50-461/97006-03g) of T.S. 5.4.1. This issue was of
increased concem to the inspectors because the Operations Department staff
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interviewed could not explain to the inspectors how the status of " Risk Basis"
design deviations was tracked. This indicated a potential problem with the
licensee's maintenance and modification program. This issue was discussed with
the licensee.
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MWR D60080, J/Ss 93A,93B,93C,93D, and 93E provided a detailed sequence of
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steps to be performed by electric maintenance mechanics to check the operation of
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the 1821F032A actuator. The inspectors identified that these J/Ss did not refer to
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procedures, written instructions, or drawings, and were therefore inconsistent with
i
the expectations for the use of MWRs as described by licensee management. After
several days of review, the licensee's maintenance staff provided the inspectors a
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copy of a completed Step 8.1.1.5 of CPS 9861.020002, "LLRT Data Sheet for
1MC009," an approved surveillance procedure, which provided step by step
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instructions to perform the test described in MWR D60080, J/Ss 93A through 93E.
(
This test had been performed as a portion of the 1821F032A modification post
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maintenance testing. The inspectors concluded that the requirements of T.S. 5.4.1
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and 10 CFR 50, Appendix B, Criterion V had been satisfied by the performance of
CPS 9861.02, step 8.1.1.5, but conveyed to plant management the concern that
the inclusion of J/Ss 93A through 93E in MWR D60080 indicated that the
programmatic guidance in CPS 1501.02 and CPS 1029.01 was either not clear or
that maintenance planners had not implemented plant managements expectations,
as described to the inspectors.
I
Status of the 1821F032A Check Valve
The licensee completed post modification testing of 1821F032A during this
reporting period. The check valve passed the tests. The inspectors noted that
1B21F032A had passed previous post-outage local leak rate test (LLRT)
surveillances but had failed subsequent as-found LLRTs. These failures were
documented in Inspection Report 50-461/96009, and indicate the need for
additional confidence in the fabrication and assembly of the valve modification.
The inspector's reWe'et of the MWR D60080 work package did not identify any
specific examples of out-of-tolerance fabrication details or improper valve assembly
techniques or conditions,
c.
Conclusions
The inspectors reviewed MWR D60080, the work authorization and coordination
4
document for the modification of the outboard check valve in the "A" Feedwater
line, and identified that it contained very poor documentation of the completion of
,
this safety-related activity. Seven examples of a violation of the Technical
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Specification for Procedures were identified.
M1.7 Installation of Drvwell Head
a.
Insoection Scene (62703)
The inspectors observed the pre-job briefing, the placement of the head, and final
local leak rate testing (LLRT).
b.
Obsersations and Findinas
The pre-job brief was considered good in that radiation protection, safety, and task
specific issues were addressed. The crew asked several useful questions during the
briefing. Following the brief, the crew divided into smaller groups for each specific
task and discussed individual assignments.
The actual work of moving and setting the head was not complicated, although
most of the work was performed in a high contamination area. Each succeeding
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task was accomplished in a deliberate and controlled manner. At one point a
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mechanic identified a minor problem with a the rigging equipment. The work was
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stopped momentarily until a replacement part was provided. The task was
accomplished without incident.
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Prior to starting the local leak rate testing of the DW head, the test operator
determined (through communications with a mechanic) that the test hose was not
I
connected properly. A mechanic had attached the hose while in the high
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contamination area performing other work. This mechanic had not been informed
of the specific hose configuration requirements. The intention to save an additional
entry into the area was defeated by poor communications. The test hose was
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reconfigured and the test proceeded in accordance with the procedure.
.
c.
Conclusions
t
Landing of the drywell head was performed in an appropriate manner. No technical
or radiological control noncompliances were identified, but poor comrnunications
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resulted in an avoidable entry into a high contamination area.
M1.8 Reactor Core Isolation Coolina (RCIC) System
a. Insoection Scone
The inspectors reviewed RCIC vendor manual requirements pertaining to preventive
maintenance (PM) and lubrication practices, then compared these requirements with
the maintenance history. Also, the inspectors reviewed the current corrective
maintenance backlog for the RCIC pump and turbine as well as any actions and
trends implemented as a result of the Maintenance Rule.
)
b. Observations and Findinas
The inspectors concluded that the current maintenance PM program for RCIC
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adequately implemented vendor manual requirements. Maintenance history
indicated that the PMs had generally been completed when the task was scheduled.
Lubrication requirements for both the pump and the turbine as stated in NSED
Standard MS-01.00, " Equipment Lubrication Standard," were consistent with the
vendore recommendations. NSED Standard MS-03.00 (Rev. 22), " Oil Sampling and
Analysis," had been updated since problems with moisture in the turbine oil were
identified in the summer of 1996 (see IR 50-461/96009).
The inspector interviewed the applicable system engineers to determine how the
system was being treated in terms of the Maintenance Rule. Plant staff stated that
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the system was in the increased monitoring a(1) status due to it not meeting
system availability goals (other goals were met). The engineers also documented
corrective actions in this area, namely to decrease unavailability by increasing
preventive maintenance and surveillance outage frequency. They had also re-
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evaluated the initial average historical unavailability of the system and found it to be
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maccurate but conservative. These actions appeared appropriate
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c.
Conclusions
The inspectors determined that the RCIC System was being maintained per the
vendors preventive maintenance recommendations. Lubrication of the pump and
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turbine was also being maintained in accordance with vendor recommendations.
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No significant problems with material condition or preventive maintenance were
identified.
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M3
Maintenance Procedures and Documentation
M3.1 Freeze Seal lasues
a.
Insnection Scoos (92700)
The inspectors reviewed data pertaining to an industry event (Licensee Event Report
from Diablo Canyon Power Plant 1) on the topic of setting two freeze seals on a
common line,
b.
Observations and Findinas
.
Maintenance Work Request (MWR) No. D60031 was initiated in April of 1995 to
document valves with leakage past their seats on the "B" Reactor Recirculation
(RR) drain line. The licensee decided to replace the leaking valves. Because the RR
loop maintenance isolation valves also leaked past their seats, a freeze seal was
,
required to isolate the drain line from the loop.
'
Field work commenced on November 11,1996, utilizing MWR D60031 and
procedure CPS 8208.01, " Freeze Seals." The MWR planned for only one freeze
seal to be installed, but did not specifically prohibit more than one. The
maintenance mechanics who installed the freeze sealinstalled a second seal
approximately six inches from the first because they were not confident the first
seal would hold. The mechanics stated that they followed all procedural
requirements for each individual freeze seal.
Subsequent to the repair, licensee engineers became aware of LER 50-275 which
described how pipe between two freeze seals was damaged due to the hydraulic
pressure generated as water between the freeze seals expanded and went from a
liquid to a solid state. The engineers walked-down plant piping systems where
similar freeze seals were known to have been established. The engineers identified
that the RR drain line area where the two freeze seals had been established under
MWR D60031 was " bubbled" by approximately .040 inch. This damage was not
visible without using calipers. The damaged pipe was subsequently replaced.
Procedure Adeouacy
The inspectors reviewed CPS 8208.01, Revision 9. The inspectors found that the
procedure provided no precautions pertaining to setting two freeze seals on one
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line, though it did discuss pressure relieving requirements in the event a freeze seal
was to be installed in the vicinity of a pump or valve. The procedure also did not
implement the Electric Power Research Institute (EPRI) Guide entitled " Freeze
Sealing (Plugging) of Piping." The EPRI guide discussed in detail the need for
properly spacing a freeze seal from another freeze seal or fixed component.
Trainina Adeouacy
The inspectors reviewed the qualifications of the two mechanics who performed
the RR drain line freeze seal activities. One of the two individuals was qualified by
virtue of a Training Waiver Form. Freeze seal specific training was covered by a
site specific lesson plan (Skill No. 082803C018, " Block A Line By Use Of Freeze
Seals"), but the training waiver was completed based on the mechanic's past
employment as a pipefitter tradesman. The justification stated on the Training
Waiver Form was " Supervisor evaluation of experience obtained at Clinton Power
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Station, craft certification letter previously submitted". This same Training Waiver
Form documented a waiver of training in operating valve lapping machines, the use
of hand signals to direct crane operators, performing flame cutting, repair / inspect
plug valves, and twelve other maintenance attributes. It should also be noted that,
according to the Supervisor of Maintenance Training, the lesson plan discussed for
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these activities has never been taught by the Nuclear Training Department. All
plant site qualifications for installation of freeze seals were based on waivers. The
inspectors considered the extensive use of training waivers an inspection Follow-up
item (50-461/97006-04) pending review of the basis for such waivers. The second
individual was not qualified, which was allowed by plant procedures as long as the
non-qualified mechanic was under the direct supervision of the' qualified mechanic.
Suoervisorv Oversicht
The inspectors determined that there was no record of supervisory oversight in the
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decision to install a second freeze seal on the RR system drain line, or the decision
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to locate that sealin close proximity to the first seal.
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Comoliance issues
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The inspectors concluded that licensee staff had installed a second freeze seal on
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the RR system drain line in a manner which resulted in damage to the reactor
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coolant system boundary. The installation of this second freeze seal was not
authorized by the MWR in use, and installation of a second freeze seal was not
adequately controlled by the freeze seal procedure, freeze seal training, and
supervisory oversight of the work. The inspectors considered this to be a case of
plant staff applying a procedure in a manner for which it was not intended rather
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than a case of a procedure inadequate for its intended purpose. Use of a procedure
which was inadequate to ensure the controlled performance of an activity affecting
quality was considered to be a Violation (50-461/97006-05) of T.S. 5.4.1.
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The inspector also reviewed the licensee's response to NRC Information Notice (IN) No. 91-41, " Potential Problems With The Use Of Freeze Seals." This IN did not -
specifically identify a concern with placing two freeze seals in proximity to each
other, but did stress the need for appropriate training and procedures when
installing freeze seals. This document was issued by the NRC in June of 1991 and
received a formal review at Clinton Power Station. The inspectors determined that
licensee staff closed their review of IN 91-41 based upon the existence of plant
procedures and training. The inspectors could not determine, based upon the
closure documentation, that a review of the procedure and training adequacy was
performed when IN 91-41 was assessed.
c. Conclusions
The licensee performed an effective follow-up on an LER from another utility, and
identified that a RR system drain line had been damaged by installation of two
freeze seals, in close proximity to each other, during the current outage. The
inspectors reviewed the installation of these freeze seals and identified that the
procedure, training, and supervisory control of this evolution was inadequate to
prevent damage to the reactor coolant system boundary. One violation of NRC
requirements was ' identified.
M3.2 Inacoropriate Test Pressure Specified for System Leakaae Tests
a.
Insoection Scone (73051,73753)
The inspectors reviewed the test conditions specified for the reactor coolant system
leakage and hydrostatic tests. The specific test procedures reviewed were:
-CPS 2800.03
Reactor Coolant System Leakage Test, Revision 13, approved
December 9,1993
-CPS 2800.03
Reactor Coolant System Leakage Test, Revision 14, approved
April 17,1995
-CPS 9059.01
Reactor Coolant System Leakage Test, Revision 2, approved
March 19,1997
b.
Qhservations and Findinas
The inspectors noted the following licensing basis commitments associated with
reactor coolant system leakage and hydrostatic test:
Figure 1.1-1 and Section 1.2.2.3.2 of the Updated Safety Analysis Report
(USAR) described the nominal operating pressure in the reactor vessel at
rated power as 1040 psia which corresponds to 1025 psig. In addition, the
inspectors noted that the licensee typically operated the plant with a
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pressure above 1000 psig in the reactor vessel steam dome when at 100
percent rated power for the period April 1995 through September 1996.
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Section 5.2.4.10 of the USAR stated that pressure-retaining Code Class 1
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component and system leakage and hydrostatic testing will be conducted in
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accordance with the requirements of Article IWB-5000 of ASME Boiler and
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Pressure Vessel Code Section XI.
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Article IWB-5221, a subsection of Article IWB-5000, of ASME Boiler and
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Pressure Vessel Code Section XI,1980 edition through Winter 1981
addenda, specified that system leakage tests be conducted at a test
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pressure not less than the nominal operating pressure associated with 100
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percent rated reactor power.
1
The test pressures specified by procedures 2800.03, revision 14, and 9059.01,
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revision 2, were based on an operating pressure of 900 psig. Both procedures
specified higher test pressures, 906 (+ 145, -0) psig and 927 (+ 103, -0) psig
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respectively, to allow for instrument error and gauge heights. The inspectors noted
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that the 906 and 926 psig pressures specified were less than the 1025 (+ 35, -0)
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paig pressure specified previously by procedure 2800.03, revision 13, and the
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operating pressure described in the USAR. The inspectors determined that the
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lower test pressure was based on the inservice test (ISI) group's interpretation that
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" nominal operating pressure" consisted of the full operating band that the reactor
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could operate at (documented by memorandum Y-99740, dated June 29,1992).
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Engineering personnel stated that the interpretation was provided to allow greater
flexibility in establishing test pressures so as to provide greater margin from the
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pr9ssure-temperature limitations established by TS. Engineering staff added that
!
this change had been requested by Operations personnel in the 1992 time frame.
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The inspectors determined that at the time the interpretation was provided, TS did
not allow increasing reactor coolant system temperature above 200* for the
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purposes of the system leakage or hydrostatic tests without undergoing a mode
change.
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The inspectors questioned the use of the lower test pressure and discussed this
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issue with licensee staff. The ISI group had established the lower limit of the
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operating band as 849 psig based on the point at which Main Steam Isolation
Valves (MSIVs) close on decreasing pressure. However, the ISI group specified
that 900 psig be used as the lower limit for testing because " continuous operation
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near the low pressure set point is not recommended because it could result in poor
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steam quality or initiation of MSIV closure." The upper limit was established as
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1045 psig based the upper o;;erating limit specified by Technical Specifications
i
(TS). The inspectors noted thnt the ISI group's interpretation of normal operating
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pressure at rated power was inconsistent with the USAR and actual system
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pressures during normal operation. Based on these observations, the inspectors
considered the ISI group's determination of " nominal operating pressure" to be
inappropriate and non-conservative.
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A reactor coolant system leakage test was performed using procedure CPS
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2800.03, revision 14, on April 20,1995. The test pressure was 964 psig. The
test was reviewed and accepted on April 28,1995, and the plant underwent an
operating cycle, May 1995 through September 1996, based on acceptance of the
test. During the sixth refueling outage, another system leakage test was performed
using procedure CPS 9059.01, revision 2, on March 23,1997. The test pressure -
was 950 psig. The March 1997 test was not accepted due to observed leakage
which required repair. The inspectors considered the performance of the April
1995 test using CPS 2800.03, revision 14, which did not include appropriate
quantitative acceptance criteria for determining the satisfactory accomplishment of
the important activity of leak checking the reactor coolant system at the nominal
operating pressure associated with rated power, to be a violation (50 461/97006-
06) of 10 CFR 50, Appendix B, Criterion V, " Procedures."
The inspectors determined that TS had been revised to permit the average reactor
coolant system temperature to be above 200* without requiring a mode change for
the purposes of system leak and hydrostatic testing. This revision was
accomplished by TS amendment 95 which was issued December 2,1994 and
became effective January 1,1995. The NRC safety evaluation for the amendment
described the reason for the change as allowing system leakage and hydrostatic
testing to be performed without exceeding pressure-temperature limitations.
Consequently, the inspectors concluded that original reason for reducing the test
pressure, to provide additional margin from pressure-temperature limits, was no
longer valid by the time the test procedure had been revised. The inspectors
considered this to be an example of poor communications between licensee
organizations contributing to the adoption of an non-conservative position.
c.
Conclusions
The inspectors identified that the licensee procedures for performing reactor coolant
system leakage tests were inadequate in that they allowed test pressures
significantly below those experienced during normal power operations. The
procedures were used during the 1995 and 1997 refueling outages. A violation of
10 CFR 50, Appendix B, Criterion V, was identified.
M8
Miscellaneous Maintenance issues
M8.X Miscellaneous Maintenance Observations
(Closed) Insoection Follow-uo item 50-461/96015-04: Flexitalic Gasket Toraue. The
item was related to the proper gasket material arid torque values to be used in the
reassembly of containment equipment drain sump check valve 1RE0388. As part
of the licensee's review of this 'mue, the vendor was contacted and verified that a
flexitalic gasket was the corra" 4pe of gasket for use in 1RE0388. However, the
torque values given within the rnointenance work request (MWR) were for use with
a corrugated type gasket. Job step seven of the MWR stated that torquing of
1RE038B should be completed in accordance with Attachment 1 of the MWR. The
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inspectors review of Attachment 1 determined that it was not clearly stated that
the torque values were only to be used with corrugated gaskets. The inspectors
concluded that the inadequate documented instruction was not a violation of NRC
requirements because the equipment drain system was not safety-related as defined
in the plant's quality assurance program or Regulatory Guide 1.33. After receiving
the proper gasket from the valve manufacturer and obtaining the proper torque
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values,1RE038B was reassembled and tested satisfactorily. The inspector follow-
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up of this item did not identify any regulatory non-compliances, and the item is
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considered to be closed.
Ill. Enu!neer*na
E1
Conduct of Engineering
E1.1
Identification of Surveillance Procedure Adeausev Problems
a.
Insoection Scone (37551. 92902)
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The inspectors monitored the licensee's response to a licensee identified problem
)
with surveillance procedures for TS required conditions.
b.
Observations and Findmas
Plant engineering personnel identified that some TS required plant conditions and
<
system capacities were incorporated directly into the facility's surveillance
.
procedures without allowance for instrumentation inaccuracies. This practice was
.
reported to be based on a mistaken assumption that the TS values had been
developed with margin for instrument inaccuracies. Once this concern was
identified by engineers who were reviewing the Standby Liquid Control System, the
licensee implemented a program for reviewing all TS referenced quantitative values
for plant conditions and system capacities to determine whether the associated
surveillance procedures were adequate. This issue did not apply to TSs for safety-
related instrumentation. Instrumentation TS values were established with
consideration of instrument inaccuracies included.
The licensee had identified several surveillance procedures inadequacies associated
with the failure to consider instrument errors at the end of the inspection period.
Each inadequacy was being documented on a CR. The inspectors planned to
perform an independent assessment of the surveillance procedure reviews and to
assess the regulatory significance of each procedure inadequacy at the completion
of the licensee's evaluations. The issue of potentially inadequate surveillance
procedures and retrospective operability determinations were considered to be an
Unresolved item (50-461/97006-07) pending inspector review of the licensee's
completed evaluations.
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c.
Conclusions
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The licensee's engineering organization identified that some TS required quantitative
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values had been incorporated directly into surveillance procedures without
consideration of instrument inaccuracies. The inspectors considered the original
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practice of using uncorrected TS values in surveillance procedures to have been
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poor, but concluded that the identification and aggressive response to this issue
1
were an example of a better questioning attitude and safety focus within
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Engineering. An unresolved item was opened pending inspector review of the
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licensee's evaluations of procedural adequacy and past operability.
0
E8
Miscellaneous Engineering lasues
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E8.1
(Closed) LER 50-461/96-018-00 and Unresolved item 50-461/96015-07: Incorrect
torous values utilized for Control Rod Drive (RD) Hydraulic Control Unit (HCU)
installation. Inspection Report Number 50-461/96015 documented a licensee
identified deficiency pertaining to the potential over-torquing (during initial
installation) of the middle and upper RD HCU retaining 3/8 inch bolts during initial
construction in 1981/1982. The inspectors concluded that the corrective actions
and analysis taken since discovery of the deficiency have been appropriate in that
all effected bolts were replaced. The inspectors reviewed the licensee's
retrospective operability determination and did not identify any non-compliances.
1
The inspectors did not identify any licensee corrective actions during the previous
two years which should have led to earlier identification of the over-torqued bolts.
The installation of over torqued 3/8 inch bolts on the RD HCUs constituted a
noncompliance with 10CFR50, Appendix B, Criterion lil, " Design Control," but
because the licensee identified and aggressively corrected this condition it is being
treated as a Non-Cited Violation 150-461/97006-08]in accordance with Section
Vll.B.1 of the NRC Enforcement Policy, NUREG-1600.
IV. Plant Sucoort
There were no significant Plant Support observations or findings during this inspection
period.
V. Manaaement Meetirigg
X1
Exit Meeting Summary
The inspectors presented the in pection results to members of licensee management on
April 14, following the conclusion of the inspection period. The licensee acknowledged
the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection
should be considered proprietary. No proprietary information was identified.
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X2
Pre-decialonal Enforcement Conference Summary
On March 4,1997, a pre-decisional enforcement conference was held at the NRC Region
lli office to discuss potential enforcement issues identified in Inspection Reports 50-
461/96011,96009, and 96014. The issues were related to weaknesses in the 50.59 and
operability determination program, deficiencies in the testing methodology and corrective
actions associated with the feedwater check valves, and the improper implementation of
design basis information into a preventive maintenance procedure which resulted in the
inoperability of safety related equipment. Slides used in the licensee's presentation at the
conference have been included as Attachment A to this report.
X3
Management Meeting Summary
)
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On March 6 7,1997, Dr. Jack Roe, Director, Division of Reactor Projects for Regions lli
and IV visited Clinton Power Station, in addition to a plant tour, Dr. Roe met with several
members of licensec management to discuss plant material condition, operations and
engineering issues, and the status of the startup readiness action plan.
X4
Pre-decisional Enforcement Conference Summary
'
On March 20,1997, a pre-decisional enforcement conference was held at the NRC Region
lli office to discuss potential enforcement issues identified in detail in NRC Inspection
Report 50-461/96012(DRS). At the conference, several members of the licensee's senior
management presented a summary of the subject events and those corrective actions
either taken or proposed. While the licensee discussed each violation in detail, the
corrective actions addressed for each event, as presented, appeared to be narrow in focus.
A copy of the handouts used during this presentation are attached to this report.
X5
Management Meeting Summary
On March 20,1997, a management meeting was held in the Region ill office to discuss
'
the status of the licensee's restart action plan. The licensee described corrective actions
taken to date. The licensee also discussed proposed methods of assessing the
effectiveness of these actions. The importance of resolving the licensee's procedure
adherence problems was discussed at some length.
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-. -
-
.
-
.
-
.
_ . .
. .
.
. _ -
_ _ . - .
. _
.
d
INSPECTION PROCEDURES USED
>
IP 40500:
Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing
Problems
IP 62703:
Maintenance Observation
IP 64704:
IP 71707:
Plant Operations
IP 73051:
Inservice inspection - Review of Program
IP 73753:
Inservice inspection
-
IP 92700:
Onsite Followup of Written Reports of Norwoutine Events at Power Reactor
Facilities
IP 92902:
Followup - Engineering
IP 92903:
Followup - Maintenance
ITEMS OPENED, CLOSED, AND DISCUSSED
,
Ooened
40-461/97006-01
Inadequate test procedure to preform a surveillance test on a
portion of the Containment Ventilation System.
,
i
j
50-461/97006-02
IFl
Unusually large range of system flows was specified in the
surveillance acceptance criteria.
50-461/97006-03a VIO
Failure to sign off job steps in the MWR at the time work was
l
performed.
50-461/97006-03b VIO
Failure to perform procedural steps as directed.
.
50-461/97006-03c VIO
Failure to adequately complete and sign for the performance of
l
work.
)
50-461/97006-03d VIO
Late entry sing-off on J/S 46, without supporting objective
j
evidence of the completion of the step.
A
50-461/97006-03e VIO
Failure to reperform and document the additional fabrication
work which changed the actuator shaft dimensions.
.
50-461/97006-03f VIO
Failure to document engineering approval to proceed with the
>
i
" Risk Basis" reassembly of 1821F032A.
I
50-461/97006-0Jg VIO
Returning 1821F032A to service as an operable flow path for
RHR prior to resolving " Risk Basis" deviations.
i
'
50-461/97006-04
IFl
Extensive use of training waivers.
24
J
_ _ . . _ _ . _ . . _ _ . . _ - _ _ _ . _ _ . . _ . _ - _ _ - _ -
. _ _ . _ _ _ . _ . . _ _ _ _ . _ .
50-461/97006-05 ' VIO
Use of an inadequate procedure to ensure the controlled
performance of an activity affecting quality.
50-461/97006-06
Procedures 2800.03, revision 14, and 9059.01, revision 2,
were inappropriate to the circumstances.
l
50-461/97006-07
The issue of potentially inadequate surveillance procedures and
l
retrospective operability determinations.
50-461/97006-08
NCV The installation of over torqued 3/8 inch bolts on the RD HCUs
constituted a noncompliance with 10CFR50, Appendix B,
Criterion 111, " Design Control."
!
!
Closed
50-461/96015-04
IFl
The item was related to the proper gasket material and torque
values to be used in the reassembly of containment equipment
,
drain sump check valve 1RE0388.
!
50-461/96-018-00 LER
Incorrect torque values utilized for Control Rod Drive (RD)
Hydraulic Control Unit (HCU installation).
50-461/96015-07
Licensee identified deficiency pertaining to the potential over-
torquing (during initialinstallation) of the middle and upper RD
HCU retraining 3/8 inch bolts during initial construction in
1981/1982.
PERSONS CONTACTED
Licensee
W. Connell, Vice President
P. Yocum, Manager - Clinton Power Station
D. Thompson, Manager - Nuclear Station Engineering Department
R. Phares, Assistant to the Vice President
D. Morris, Director - Radiation Protection
A. Mueller, Assistant Plant Manager - Maintenance
M. Lyon, Assistant Plant Manger - Operations
I
.
I
25
l
l
. - - + .
_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
UST OF ACRONYMS
ATM
Automatic Trip Module
Division of Reactor Projects
Engineering Change Notice
Hydraulic Control Units
Main Steam isolatico Valve
MWRs
Maintenance Work floquest
POR
Public Document Room
Preventative Maintenance
Reactor Core isolation Cooling
Reactor Recirculation
Reactor preasure valve
Shift Supervisors
TS
Technical Specification
Updated Safety Analysis Report
26
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REFER TO E02-1 AF04-017 FOR THE SETPOINT.
-
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lilinois Power Company
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Clinton Power Station
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Enforcement Conference Presentation
l
March 4,1997
.
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4
AGENDA
.
,
'
Introduction
W. Connell
l
Vice President -
Clinton Power Station
CPS 50.59 Improvement Plan
P. Telthorst
Director-Licensing
'
50.59 Evaluations and Corrective Action
D. Thompson
i
Manager-Engineering
,
Concluding Remarks
W. Connell
Vice President -
!
Clinton Power Station
,
.
.
.
. -
.
.
.
. .
-
. .
.
I
- - - _
--
-_-_--
--
- - - - - - - - - - - - - - -
!
!
l
CLINTON POWER STATION 50.59 IMPROVEMENT PLAN
i
i
i
1
j
Needs identified by:
NRC Inspections
.
.
I
NSED Independent Assessment
.
-
.
'
CPS Reviews
.
i
improvement Areas include:
Ability to Recognize an Activity as a Test or Activity Not Described in the
e
!
.
!
Ability to Document Justification Why a Change Does Not Require a Safety
.
Evaluation
,
i
j
Review of Licensing Basis Documentation
e
i
- , _ . . - . _ . . . . _ _ . _ . . , _ _ . . . . _ _ _ - . - - - _ . _ . . , _ _ , - - , , _ , , _ , , . _ , - . . , ~ , . . - _ . - - _ . . . . - _ . _ , .
. _ - - _ _ . ,
-
,
!.
_. ---
- - - - - - - - _ -
--- --- - .
!
!
t'
CLINTON POWER STATION 50.59 IMPROVEMENT PLAN JCont.?
-
.
l
CPS Response
!
Review of Engineering Changes implemented in RF-6
.
Training
.
1
CPS 50.59 Action Plan
.
Revision of CPS Procedure 1005.06
j
-
l
Establishment of Core Group of Reviewers
-
i
Enhanced Training for Core Reviewers
'
-
!
Sampling of 50.59 Screenings
j
-
Long-Term improvement Plan
l
.
i
!
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- - - , - . , , - - - ,
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OVERVIEW OF ENGl_NEERING PRESENTATION
i
!
.
Corrective Action
i
.
4
4
-
,
50.59 Evaluations
.
,
.
.
.
i
4
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CORRECTIVE ACTION
i
!
i
!
Cathodic Protection System (96011-05b):
!
1995 Survey to Review Adequacy of System
!
.
Certain Piping Did Not Meet the Requirements of the CPS USAR
.
,
l
Safety Evaluation Not Performed
.
i
Root Cause for Cathodic Protection:
Lack of Sensitivity and Understanding of USAR Requirements
.
.
!
Corrective Actions for Cathodic Protection:
!
l
Performance of Safety Evaluation
.
i
Programmatic Improvements to CPS 50.59 Process
l
.
i
!
i
j
l
j
.
.
.--
- - . - - - .
.
.
_ --
-
l
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CORRECTIVE ACTION (Cont.)
i
1993 VC Chiller Auto Start Event (96001-07b):
l
l
1993 Integrated Testing
.
i
1993 Engineering Work Request
.
l
RF-5 Sunteillances
.
RF-6 Surveillances
l
.
l!
Root Causes for VC Chiller Auto Start:
i
e
inadequate Operability Determination
l
Lack of Formal Operability Program
.
Corrective Actions for VC Chiller Auto Start:
!
Calibration of Time Delay Sequencing Relay
!
.
l
Comprehensive Review of Operability Determinations
.
!
Verification by Independent Contractor
.
j
Creation of Formal Operability Determinations
.
'
Implementation of Design Change Issue
.
Verification of. Existing Preventive Maintenance Tasks or Writing New Ones
.
.
Review of Other Diesel Generator Sequencing
.
-
.
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- - - _ - - _ _ _ - - _ _ _ - _ _ - _ _
- _ - - _ .
d
.
50.59 EVALUATIONS
Annunciator Response Books Maintained on Control Panels (96011-05c):
Purpose and Use of Books
.
Root Cause:
Lack of Sensitivity to Potential Seismic impact
.
Corrective Action:
Books Removed From Panel Tops in MCR and Simulator
.
50.59 Improvement Plan
.
. --- _
__
. - -
- _ - _ - - _ _ -
.
50.59 EVALUATIONS (Cont.1
4
i
FC Pump Inlet Valves Contrary to USAR (96011-05d):
.
Update of CPS Procedure 3317.01
.
-
'
USAR Figure
.
Root Cause:
.
Lack of Sensitivity to Safety Evaluation Requirement
.
Corrective Action:
,
Procedure Change Prior to Startup
.
I
50.59 Improvement Plan
.
1
I
I
!
4
4
- - - - - - - -
- - -
-
- .
- - - - - . - - - -
- - - - .
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I
,
Ineffective Corrective Action (96009-08a):
!
Purpose and Description of Valves
.
Maintenance and Testing History
.
RF-6 As-Found Test Failure
.
Root Causes:
Failure to Take Effective Action
.
4
' Failure to Pursue Alternative System Design
.
i
Corrective Action:
.
Modify the Valves and Actuators
.
32A and 32B Valves Returned to Operable Status
e
Explore Alternate System Design Solution
.
4
m.
.
--
,.
-
1
i
FEEDWATER CHECK VALVES (Cont.)
.
.
Failure to Follow Testina Procedure (96009-08b):
i
ANSI Standard 56.8-1994
.
.
I
Corrective Maintenance
.
!
Testing Performed
.
Root Causes:
Lack of Familiarity with Design
'
.
' Original Design-does not Allow Effective Draining
.
Corrective Action:
Modification of Design
.
Revision of Test Procedures
>
.
Review of Oth.er Type C Tested Containment Penetrations
.
,
1
- _ , . . -_.
_ . , - . . _ _ - . , . . - _ . , . _ . . .
..
. _ . . . _ , . . _ , _ . . . . . , _ . -
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-
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.
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-
!
!
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Description of Violations (96014-01a and 1b):
!
Relay Setpoint Information Not Correctly Translated into Procedures
.
!
Discrepancy Between As-Found and As-Left Relay Setpoints Not
l
.
i
Questioned
,!
Identified by. Licensee during Surveillance
.
!
Root Causes:
i
Misinterpretation of Setpoint Data
.
l
i
Lack of Rigorin Engineering Activities
"
.
~
Lack of a Questioning Attitude
.
i
e
i
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.
.
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INOPERABILITY OF EMERGENCY DIESEL GENERATOR JCont.)
i
i
Corrective Actions:
!
Relay Replaced and Calibrated to Correct Setpoint
!
.
i
i
i
Integrated Surveillance Test Satisfactorily Performed
.
i
)
improved Drawing
=
!
,
j
Verification of Remaining Setpoints
.
Additional Verification of Electrical Relays
.
4
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OVERVIEW OF RADIATION PROTECTION
,
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Record of Strong Performance
i'
1
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Outage Dose Reduction
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SALP 1 Rating
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Department Philosophy
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APPARENT VIOLATIONS
Radworker Performance
Drywell RSWP
Waste Sludge Handling
Insulation Removal
_ _ - - -
_ ---
i
l
RADWORKER PERFORMANCE
!
,
i
CPS Identified Examples of Unacceptable Worker Performance
!
!
Root Cause:
i
Lack of Strong Line Accountability
Contributina Causes:
Rationalizing Away RP Requirements
Station Radon Problem
Worker Performance Problems Not Sufficiently Visible to Line Organization
i
i
!
RADWORKER PERFORMANCE CORRECTIVE ACTIONS
1
,
immediate Corrective Action:
l
!
l
Expectations on RP Requirements
!
Replacement of Radiological Deficiency Reports with Condition Reports
!
1
!
Increased Monitoring of Radworker Practices
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_ . _ , _ _ _ . . . _ . . , _ _ _ _ , _ - _ , , _ . . _ _ - _ , _ . _ . _ _ . _
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- _ - _ _ _
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RADWORKER PERFORMANCE CORRECTIVE ACTIONS
,
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Lona Term Corrective Action:
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Continuance of the Use of Condition Reports
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!
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Implementation of Remedial Action Approach
!
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Plan to Address Radon Problem
i
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Improvement Plan for Radworker Knowledge of Radiological Requirements
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and Practices
!
Department Self Assessment
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-
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IMPLEMENTATION OF
!
DRYWELL RADIATION SAFETY WORK PLAN (RSWPD
,
i
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Description and Purpose of RSWP
.
i
Outage Organizations
.
i
Summary of Event
Safety Significance
i
Root Causes:
!
Incomplete Understanding of RSWP
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Ineffective Communication
'
Lack of Clarity in RSWP
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RSWP CORRECTIVE ACTIONS
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immediate Corrective Action:
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Radiation Operations Written Communications
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Hold on Further RSWP Suspensions
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Revision of RSWP
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Reinforce RSWP with Affected Personnel
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Written Direction and Clarification of RPSS Authority
,
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Lona Term Corrective Action:
.
Enhanced Briefing of RSWP Prior to Next Outage
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Revision CPS Procedure on RSWPs
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,.. ___... - _. _ _ .___ _ _ ._. . _ . _ _ _ _ __. _ ,_ _ . _ _ _ _
.
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WASTE SLUDGE SLUICING
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Description of Sluicina Operation
!
Problems identified:
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Pre-job Briefing Lacked Discussion of Contingencies
l
.
l
Supervision Was Not Notified of Problem
Radiation Protection Technician Did Not Maintain Oversight Role
Procedures Did Not Cover Hose Blockage Actions
Vendor Procedure Did Not Address CPS Configuration
,
Lack of RP Review of Vendor Procedure
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Lack of Understanding of Vent Path by Vendor Representative
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Waste Sludne Sluicina (continued)
Root Causes:
Ineffective Management of the Process
inadequate Equipment Configuration
Equipment and Procedure Did Not Address Potential Problems inherent to
Sluicing
i
Lack of RPT Oversight
Lack of a Planned Approach
!
i
-
.
Waste Sludge Sluicina (continued)
!
Corrective Actions:
Error Prevention Training
i
Hold Placed On Use of Pump and Sluicing Activities Pending Further
Investigation and Completion of Corrective Action
Revision of Vendor Procedures and Review By Radiation Protection
Department
Procurement of Wand Matching Procedure Configuration
Revision of Vendor Equipment Drawing Applicable to CPS
I
_ - - - -
- - - - - _ - - - - - - - - - - - - - -
-- ---
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INSULATION REMOVAL
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Description of Respirator Evaluation and RWP Pass Programs
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Description of incident
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Results of CPS Investigation of incident:
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Contamination Levels Exceeded Allowable Levels
!
.
Proper Precautions to Prevent Spread of Contamination Not Taken
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Wetting of Insulation Nor Performed
!
!
Inadequate Radiation Work Permit and RWP Pass Card
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RPT Monitoring of Insulation Removal Insufficient to Prevent Exceeding
Contamination Limit Set
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Inadequate Understanding of Ventilation System Status
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INSULATION REMOVAL Jcontinued)
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,
Root Cause:
Inadequate Implementation of Engineering Controls
!
,
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Immediate Corrective Action:
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Decontamination of Affected Personnel and Areas
,
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Performance of Diagnostic Whole Body Counts
Specific Event Training for RP Technicians
,
Long Term Correction Action:
Revision of Radiological Systems Lesson Plan
i
Revision of CPS Procedure on Radiological Job Coverage
i
implementation of Respirator Evaluation Engineering Control
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Director - Plant
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Radiation and
Outage Manager
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Chemistry
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Assistant Director- Plant
Radiation Protection i
Shift Outage Manager
Radiation Protection Manager
(SOM)
(RPM)
Supervisor- Radiological
Operations
Shift Work Coordinator
Drywell Coordinator
Drywell Radiation
Protection Shift
Supenrisor (RPSS)
Drywell Control Point
Lead Technician
Functional Responsibility
Communication Lines
Drywell Radiation
Protection Technicans
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Simplified Schematic
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