ML20247D296
| ML20247D296 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 05/05/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20247D285 | List: |
| References | |
| 50-266-98-06, 50-266-98-6, 50-301-98-06, 50-301-98-6, NUDOCS 9805140331 | |
| Download: ML20247D296 (30) | |
See also: IR 05000266/1998006
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U.S. NUCLEAR REGULATORY COMMISSION
REGIONlli
Docket Nos: 50-266;50-301
Report No: 50-266/98006(DRP); 50-301/98006(DRP)
Licensee: Wisconsin Electric Power Company
Facility: Point Beach Nuclear Power Plant, Units 1 & 2
Location: 6612 Nuclear Road
Two Rivers, WI 54241-9516
Dates: fiarch 3 through April 13,1998
Inspectors: F. Brown, Senior Resident inspector
P. Louden, Resident inspector
P. Simpson, Resident inspector
Approved by: J. W. McCormick-Barger, Cliief
Reactor Projects Branch 7
9805140331 980505
G ADOCK 05000266
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EXECUTIVE SUMMARY
Point Beach Nuclear Plant, Units 1 & 2
NRC Inspection Report No. 50-266/98006(DRP); 50-301/98006(DRP)
This inspection includM aspects of licensee operations, engineering, maintenance, and plant
support. The report covers a six-week inspection period by the resident inspectors.
Operations
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Operations personnel involved with the restart of the Unit 2 reactor exercised good
control of reactivity changes. Clear, consistent communications were used by operators.
(Section 01.1)
. A reactor operator who was "at the controls" for a unit that was shut down and defueled,
left the authorized surveillance area for a short period of time without being appropriately
relieved by another reactor operator. This action was contrary to the requirements of the
licensee procedure for the conduct of operations and was a violation of Criterion V,
" Instructions, Procedures, and Drawings," of 10 CFR Part 50, Appendix B. (Section 01.2)
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Operators responded appropriately when the second stage seal of an idle reactor coolant
pump partially opened. Planning of the pump restart and communications and procedure
adherence during the restart were appropriate and effective. (Section O2.1)
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The use of tape to cover the bearing grease port of the residual heat removal pump motor
instead of the vendor-designed cover reflected an acceptance of substandard conditions
by auxiliary operators. (Section O2.2)
Maintenance
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Main control board wire separation work was conducted in a professional and thorough
manner. All work observed was performed with the appropriate work order plan present
and in active use. (Section M1.1)
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Maintenance and health physics organizations were not effectively prepared to perform
the lower intemals lift based on planning meetings conducted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the
initiation of work. Early in the evolution, maintenance workers failed to follow procedures
resulting in a violation of Technical Specification 15.6.8.1. Laterin the evolution, the
maintenance organization displayed better control of the activity, and the lower intemals
were moved without incident. (Section M1.2)
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Many observed maintenance activities were completed in accordance with requirements
specified in administrative and work control procedures. However, ceases were noted
where administrative requirements were not being implemented. Some of the corrective
actions for these issues were narrowly focused, and the effort to address the
inconsistencies in application of administrative requirements within the maintenance
department was not an integrated effort. (Section M1.3)
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. Maintenance and operations department freeze seal pre-evolution briefings held on
March 17,1998, were thorough and covered command and control responsibilities,
expected communication standards, and contingencies. Teamwork between different
disciplines was evident and participants displayed a good questioning attitude.
(Section M1.5)
Enaineerina
. A ventilation control panel in an emergency diesel generator room was misclassified as
nonsafety-related. The licensee's initial corrective actions did not include determining if
operability of the system had been challenged while the component was incorrectly
classified as being nonsafety-related. (Section E1.1)
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The licensee identified and corrected two cases where valves between seismically
l qualified piping systems and non-qualified piping systems were not maintained in a
I closed position as required by the Final Safety Analysis Report. (Section E1.2)
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. Offsite, corporate office-based engineering personnel working on a corrective action
commitment initiative to assess the adequacy of a separation of seismically qualified and
! non-qualified piping systems were performing analyses and taking credit for components
to function in a manner that may not have previously been considered in the design basis.
The engineers had not evaluated whether such reliance might constitute a design basis
j change. Additionally, onsite licensee personnel performing concurrent and interrelated
- corrective action initiatives had not been informed of the potential design engineering
activities that could have affected the results of these other initiatives. (Section E1.3)
. The inspectors concluded that the 125-Volt direct current (Vde) system was capable of
l meeting design basis functions. However, the failure to maintain an up-to-date battery
loading calculation was considered a violation of 10 CFR Part 50, Appendix B,
Criterion Ill, " Design Control." (Section E3.1)
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. The reactor engineering organization did not provide accurate critical rats position data to
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operations personnel during an initial attempt to startup Unit 2. The problems revealed
during the startup were considered additional examples of reactor engineering
performance concems which were the subject of a Notice of Violation from Inspection
l Report No. 50-266/98003(DRP); 50-301/98003(DRP). (Section E3.2)
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. The practice of duty technical advisors (DTAs) serving two consecutive 24-hour watches
was not consistent with the intent of program procedures and raised questions regarding
the DTA's fitness-for-duty. Although, no specific performance issues were identified as a
result of the DTA standing consecutive watches, licensee management immediately
revised expectations regarding this practice to preclude potential fitness-for-duty issues.
(Section E6.1)
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Plant Support
. Perronnel exposures during the Unit i refueling outage were meeting established
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licensee goals. The number of personnel contamination events was higher than
l anticipated; however, most of the events were minor shoe contaminations. The health
physics manager initiated a review of the causes for the higher than anticipated number
of personnel contamination events. None of the events resulted in significant exposure of
! personnel. (Section R1.1)
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Report Details
Summary of Plant Status
During this inspection period, Unit 1 was shutdown in a continuation of the Cycle 24 refueling
outage. ' Unit 2 was shutdown on March 5,1998, in accordance with Technical Specification
(T/S) 15.3.0., because the compenent cooling water (CCW) system was declared inoperable.
Detailed engineering analysis subsequently determined that the CCW system was operable.-
Unit 2 was restarted on March 28,1998, and operated at 100 percent power for the remainder of
the inspection period.
Inspection Focus
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j During this inspection period, the inspectors focused on conduct of plant operations, continued a
vertical slice review of the 125-volt direct current (Vde) system, and completed routine inspection
activities.
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1. Operations
01 Conduct of Operations
l 01.1 Unit 2 Reactor Startuo (Inspection Procedure (IP) 71707)
During the restart of the Unit 2 reactor on March 28,1998, problems encountered during
the attempt to make the reactor critical resulted in the licensee suspending the criticality
l attempt. The reactor was made critical later the same day following a review of the
earlier problems and a recalculation of the estimated critical rate position. Problems
associated with the initial estimated critical rate position calculation are discussed in
Section E3.2 of this report. Operations personnelinvolved with the restart of the Unit 2
reactor exercised good command and control of reactivity changes and used clear,
consistent three-way communications.
01.2 Unit Operator Left the Authorized "At the Controls" Area
a. Inspection Scope (IP 71707)
The inspectors reviewed the circumstances regarding the failure of an onshift reactor
operator (control operator (CO)) to remain within authorized surveillance areas in the
- control room.
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l b. Observations and Findingg
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The inspectors were in the control room monitoring a Unit 2 non-routine activity on ;
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March 14,1998. At approximately 9:00 p.m., the inspectors noted that the Unit 1 CO was
l not in an authorized surveillance area for Unit 1, which was defueled at the time. Shortly
i. thereafter, the Unit 1 CO reentered the authorized surveillance area from the control room
back panel area. The CO was absent from the authorized area for about one minute.
The T/S minimem manning requirements were satisfied during the CO's absence; i
however, Operations Manual (OM) 1.1, " Conduct of Plant Operations," Revision 1,
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Attachment 2, Paragraph 2.3, required the reactor operator "at the controls" to remain in
the authorized area unless relieved. The inspectors discussed the CO's absence from
the authorized area with the duty operating supervisor (DOS, a senior reactor operator).
The DOS stated that the CO had been released from the authorized area for a short
period of time, and that this was acceptable because Unit 1 was defueled. The
inspectors pointed out that OM1.1 allowed no exceptions. This issue was further
discussed with the operations manager, who acknowledged that OM 1.1 required the unit
CO to remain in the authorized area under all fuel loading conditions. The failure of the
Unit 1 CO to remain in the authorized area was a violation (VIO 50-266/98006-01(DRP))
of 10 CFR Part 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings,"
which requires that activities affecting quality be performed in accordance with
procedures. Condition Report (CR) 98-1075 was written to document this event, and the
operations manager sent all operators an electronic memorandum which reiterated the
requirements of OM 1.1 for an operator "at the controls."
The inspectors also identified a discrepancy in OM 1.1. Figure 1 and Section 2.8 of
Attachment 1 differed concoming the control room area the DOS was to occupy. The
licensee's practice was to allow the DOS to sit on a raised platform in the control room,
which was consistent with Section 2.8 but was not allowed by Figure 1. The inspectors
identified the discrepancy to the duty shift superintendent (DSS, a senior reactor
operator). The DSS stated the discrepancy had already been identified by operations
personnel via the procedure change process about six months earlier but was not yet
corrected. The DSS wrote a CR to document the discrepancy on March 19,1998. This
discrepancy was corrected the same day with a temporary procedure change.
During this inspection period, the inspectors noted that the operations department had a
significant number of outstanding procedure change requests and was in the process of
upgrading operations procedures. Licensee management indicated that priorities were
set to accomplish the procedure upgrade work within existing resource constraints. The
correction of OM 1.1, identified six months earlier by operators, was not high in the
priority scheme. The inspectors commented to operations management that procedural
adherence and operator identification of needed procedure changes may be adversely
affected given the large backlog which impacted the timeliness of processing procedure
changes. However, the inspectors noted that progress was being made in upgrading
operations department procedures overall.
c. Conclusions
The inspectors concluded that the CO who left the "at the controls" area for a bnef time
on March 14,1998, without obtaining an appropriate relief, was not performing duties in
accordance with OM 1.1. This was considered a violation of 10 CFR Part 50,
Appendix B. The inspectors also identified a discrepancy in OM 1.1, which the licensee
subsequently corrected. The procedure upgrade program contained a substantial
backlog of identified changes that needed to be made; however, a prioritization list was
being followed and some progress was being made.
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02 Operational Status of Facilities and Equipment
O2.1 Reactor Coolant Pumo (RCP) Seal Leakaae
a. (nspection Scope (IP 71707)
The inspectors reviewed the licensee's response to excessive leakage from the
Unit 2 "B" RCP (2P-1B) second stage seal.
b. Observations and Findings
On March 10,1998, operators noted excessive flow through the second stage seal of
RCP 2P-1B. Unit 2 was in cold shutdown and reactor coolant pressure was 300 pounds
per square inch gauge (psig) with the "A" RCP (2P-1 A) operating and the "B" RCP (2P-
1B) idle. Upon the discovery of the excessive seal leakage, the operating crew entered ;
Abnormal Operating Procedure 18. "RCP Malfunction," Revision 8. In accordance with i
that procedure, pump 2P-1 A was secured, the reactor was depressurized to about i
50 psig, and tne RCP seal water retum valves were closed. These actions terminated ,
the excessive flow. The inspectors noted thet the operating crew referenced the j
appropriate T/S for reactor coolant system leakage. The licensee wrote a condition report I
to document the event. l
The licensee formed a multi-disciplinary team to assess the condition of the 2P-1B seal
package and concluded that the second stage seal had partially opened, but had not
failed. A temporary change was made to Operating Procedure (OP) 3C, " Hot Shutdown 1
to Cold Shutdown," Revision 69, to provide instructions for starting 2P-1B to allow for
further evaluation of the seal's condition. The inspectors reviewed the change to OP 3C
and the referenced sections of OP 4B, "RCP Operation," Revision 34, and concluded that
the changes were appropriate for the circumstances. An operating crew started
RCP 2P-1B without incident on March 14,1998. The second stage seal reseated during
the pump start. The inspectors observed appropriate and effective communications,
planning, and performance of pump start activities in the control room during this
evolution.
c. Conclusions
Operators responded appropriately when the second stage seal of an idle RCP partially
l opened. Planning of the pump restart and communications and procedure adherence
during the restart were appropriate and effective.
O2.2 Residual Heat Removal (RHR) Pump Motor Grease (IP 71707)
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During a routine walkdown of safety-related systems, the inspectors identified a ,
discrepancy in the amount of grease present on the outboard bearings of the l
l four RHR pump motors. The amount ranged from grease fully covering the bearings to
being hardly visible. The inspectors also noted that tape was used to cover the
Unit 1 RHR "A" pump motor outboard bearing grease port in lieu of the vendor-designed i
cover. When notified of the findings, the component engineers investigated the situation
and informally determined that the pumps were still operable. Operations personnel
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wrote a condition report regarding the matter and requested a formal operability I
determination (OD). The licensee concluded in the OD that the pumps remained
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The inspectors concluded that the bearing grease levels were not an operability concem.
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L However, the use of tape to cover the bearing grease port of the safety-related
RHR pump instead of the vendor-designed cover reflected auxiliary operator acceptance
of substandard conditions.
O3 Operations Procedures and Documentation ;
O3.1 Update on Station Wide Procedure Uparade Proaram (IPs 71707. 62707. and 37551)
In a previous inspection report (No. 50-266/97020(DRP); 50-301/97020(DRP)), the
inspectors opened an inspection follow up item (IFI) to evaluate the licensee's ongoing
procedure upgrade program and verify that: j
. upper tier administrative procedures for procedure adherence and procedural
control of activities were consistent wia the current licensing basis and NRC
guidance,
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- the methods for establishing procedural controis were commensurate with J
L licensee staff training and supervisory oversight such that activities affecting i
safety were performed in a controlled manner and with predictable results, and
+ the licensee's process for assuring that work plans were not inappropriately used
to circumvent procedural change requirements were adequate. ]
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Over the past six months, the inspectors have identified several instances where
procedural controls were either inadequate for the circumstances or were not adhered to l
by licensee personnel. Sections 01.2, M1.2, M1.3, and E3.2 of this report discuss other i
examples of procedural problems still evident at the station. .
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Notwithstanding these problems, the inspectors have noted an increased sensitivity to
procedural quality issues and some progress in upgrading procedures. Additionally,
insufficient time has passed to determine the effect of the licensee' procedure upgrade
project. Therefore, the inspectors will leave IFl 50-266/97020-02(DRP);
50-301/97020-02(DRP) open for an additional six-month period to track the programmatic
aspects of procedure content, use, and adherence.
07 Quality Assurance in Operations
07.1 Operations Quality Assurance Audii(IP 71707)
The inspectors attended a quality assurance department audit exit on March 13,1998.
The audit focused on operations department administrative controls and operator 4
performance. The meeting was well attended by operations department personnel and
plant management. Operations management was receptive to the auditors' findings. The
inspectors noted that most of the findings from the audit were more administrative in
nature rather than performance-based. The inspectors subsequently reviewed the issued
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audit report (No. A-P-98-03) and verified that the issues discussed at the exit meeting
were consistent with those documented in the report.
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08 Miscellaneous Operations issues
08.1 (Closed) Licensee Event Report (LER) 50-266/98004: 50-301/98004: Resc.or coolant
pump lube oil collection system design nonconformance with Appendix R, Section 111.0. 1
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This issue was discussed and dispositioned in accordance with the NRC Enforcement
Policy in inspection Report No. 50-266/98003(DRP); 50-301/98003(DRP), Section O2.1.
No further action is necessary regarding this matter.
08.2 (Closed) LER 50-301/98002: Reactor coolant pump component cooling water retum line I
check valve found seriously degraded. The CCW system containment retum check
valve (2CC-745) was radiographer and found in the open position. This valve provides a
redundant means for preventing loss of CCW fluid in the event of a failure of a CCW pipe j
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inside the containment. The licensee rebuilt the intamals of the valve and the repairs !
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were deemed to be adequate. The inspectors had no further questions regarding this I
matter. I
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II. Mainte_n_ance
M1 Conduct of Maintenance
M1.1 Main Control Board Wire Separation Maintenance Activities
a. [nLspection
r Scope (IP 62707)
The inspectors observed and reviewed the following maintenance activities which were
part of the corrective actions to resolve discrepancies between redundant safety-relsted
equipment:
. Work Order Plan 9705320, " Sleeve / Wrap Cables For Circuit 1 A-06 Bus l
Voltmeter," and
. Work Order Plan 9705324, " Sleeve / Wrap Cables For Circuit Supply
Breaker 1 A52-77 to Bus 1 A-04."
The planned activities included separating and sleeving electrical wires in some control
room panels associated with control and indication circuits for the Class 1E electrical
buses and power sources,
b. Observations and Findinos
The scope of the planned cable separation work required entry into the T/S 15.3.7.B.1.g.
limiting condition for operation (LCO) for Unit 2 (Unit 1 was defueled at the time) since the
' B" train of the 4.16-kilovolt bus safeguards switchgear (Bus 1 A-06) did not have its
emergency power source available because of the protective tagout boundary. The
inspectors verified that the appropriate T/S LCOs had been entered for the plant
conditions and scope of planued work.
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The inspectors noted good coordination and communication between the work control
center, control room, and maintenance personnel. The job supervisor briefed the control
room personnel on the specifics of each work order plan and walked through each
package with the maintenance crew doing the work. Quality control personnel were
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properly verified and signed off.
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l The maintenance crews used self-verification checks in the cramped and sensitive work
j environment. Workers also displayed good questioning attitudes during the course of the
i work. Worker-identified discrepancies in work plans were called to the attention of
l maintenance supervision and corrected appropriately.
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The inspectors concluded that the control board wire separation work was performed in a ,
professional and thorough manner, All work observed was performed with the l
appropriate work order plan present and being appropriately referenced. j
! M1.2 Unit 1 Reactor Vessel Lower intemals Lift
a. Inspection Scope (IP 62707)
The inspectors observed the planning and execution of the Unit i reactor vessel (RV) i
lower intemals lift.
b. Obser<ations and Findinas -
The inspectors attended a work preparation briefing on March 3,1998, which was held to
discuss the various aspects of the lower intamals lifting evolution. Work group
l responsibilities were identified and Routine Maintenance Procedure (RMP) 9053, "RV
Intemals Removal and Installation," Revision 1, was reviewed. Health physics ,
considerations were discussed; however, radiation work permits had not been completed. '
The inspectors made the following observations regarding the briefing:
. Contrary to Nuclear Procedure 1.2.6, " Infrequently Performed Tests and
Evolutions," Revision 4, the work activity was not categorized as an infrequently
performed test or evolution. This condition was subsequently corrected prior to
the beginning of work.
. Health physics information discussed was not complete nor fully evaluated prior
to the planning meeting. For example, the radiation work permits had not been
prepared.
l . Overall, discussions at the briefing (held 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before the initiation of work)
indicated that many aspects of the job had not been thoroughly evaluated.
The initial attempt to lift the lower intemals was performed under the direction of a
maintenance supervisor. A senior manager was present in the containment to provide
oversight. The inspectors observed that the maintenance crew attached the containment
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polar crane hook to the "intemals lift rig," a special lifting platform, without inserting a load
cell between the hook and lift rig, as required by Step 7.3.7.f. of the RMP. The purpose
of the load cell was to provide early indication of binding during crane load vertical
movement. The inspectors asked the maintenance supervisor why this step had not
been completed. The supervisor stated that the lift rig was to be moved to the other side
of the containment and the load cell would then be installed. This action would have
been acceptable, since the procedure specifically only prevented the lifting rig from being
positioned above the reactor vessel without the installed load cell. However, the crew
l proceeded with positioning the lift rig above the RV with the intent of lowering the lift rig
into place on the RV and then installing the load cell. Positioning the intemals lift rig
above the reactor vessel without the load cell installed was a violation
(VIO 50-266/98006-02(DRP)) of T/S 15.6.81 for failure to follow procedures. Prior to the
lift rig being lowered onto the RV, the senior manager in containment recognized that the
procedural requirements were being violated and stopped work. The action was
documented on CR 98-0831. A temporary change was made to RMP 9053 to allow
installation of the load cell after the lift rig was landed on the RV While the procedure
change was being processed, involved maintenance personnel stated that the original
procedure had been ir;dequate. The inspectors concluded that the steps in the original
procedure could have been performed as written. Additionally, the inspectors noted that
maintenance staff had ample opportunity during the pre-job briefing to decide how to
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perform the steps as written or to identify attemate ways to perform the lift, and make any
necessary procedural changes.
The maintenance crew attempted to lower the lift rig onto the RV after the temporary
change was processed for the RMP. The three guide bushings on the lift rig were not
proper 1y aligned with the three guide studs in the RV flange, and at least one bushing was
observed to be resting on the corresponding guide tube. The full weight of the lift rig
appeared to be placed on the guide bushings which were resting on top of the guide
studs. The crane operator could not quickly identify the misalignment because the load
cell was not present to indicate reduced weight on the crane as the crane hook was
lowered. The lift rig was raised off the guide studs and rotated into proper alignment. On
the second attempt at lowering the !ift rig, the mounting plates for two of the guide
bushings were found to have been knocked out of alignment to the extent that the guide
bushings would no longer slide down the guide studs. The lift rig was transferred back to
a laydown area, and the guide bushing mounting plates were realigned. Lifting
operations were suspended to allow for a shift change of personnel.
A pre-job briefing was conducted for the on-coming shift personnel. Overall, the briefing
was conducted well. The maintenance supervisor in charge of the evolution displayed a
clear understanding of the task and clearly outlined roles and responsibilities of the work ,
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crew members. During the conduct of the intemals lift, the maintenance supervisor
maintained a " big picture" oversight of the activity. Having noted the procedural
compliance problems during the previous shift, the work crew leader was deliberate in
taking actions and frequently referenced the RMP to ensure steps were appropriately
completed. The lift was conducted very methodically and was controlled well. The lower
intemals were placed in the storage area of the Unit 1 cavity area without incident.
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c. Conclusions
The inspectors concluded that the maintenance and health physics organizations were
not adequately prepared to perform the lower intemals lift based on planning meetings
conducted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the initiation of work. In addition, early in the evolution, the
inspectors identified a failure of maintenance workers to follow procedures as required by
T/S. This was considered a violation. Later in the evolution, the maintenance
organization displayed better control of the activity and the lower internals were moved
without incident.
M1.3 Inconsistent Application of Administrative Controls in Maintenance
a. Inspection Scope (IPs 62707 and 40500)
The inspectors assessed the maintenance department's implementation of administrative
controls, including procedure adherence.
b. Observations and Findinas
Maintenance personnel were observed to be performing many maintenance activities in
full compliance with procedural and other administrative controls. However, the failure to
utilize the procedural controls in place during the lower intemals lift, and the failure to
effectively utilize the pre-job brief to ensure the appropriateness of the planned method of )
performing work during the lower intemals lift, described in Section M1.2 above, were
indicative of inconsistencies in the maintenance department's application of standard
administrative controls. The inspectors identified two other minor discrepancies in the
application of administrative controls by the maintenance department during this period.
These conditions were discussed with licensee staff and were corrected under
CR 98-0917 and CR 98-1168. Additional,' unrelated examples of inconsistent application
of administrative controls were identified by the licensee, and were documented in
CR 98-1369 ar~i CR 98-1463. Similar issues were discussed in Section M2.2 of
IR No. 50-266i98003(DRP); 50-301/98003(DRP).
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Specific corrective actions were taken for each identified discrepancy, but the inspectors
noted that there was no broad-based initiative to address the observed discrepancies. j
Additionally, some of the corrective actions were narrowly focused. For instance, the only
corrective action documented for the RV lower intemals lift procedure violation
(CR 98-0831) was a permanent change to the procedure to add greater flexibility in the
performance of work steps. This did not appear to address all of the performance issues
discussed in Section M1.2 above. This concem was discussed with the maintenance
- . manager, who indicated that the performance discrepancies were not pervasive, and that
i various initiatives were in place to improve the performance of maintenance activities.
The maintenance manager further stated that a coordinated effort to address both long-
!. term corrective actions and interim actions within the department was worth
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c. Conclusions
Many of the maintenance activities observed were completed in accordance with
requirements specified in administrative and work control procedures. However, the
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inspectors noted cases where administrative requirements were not being implemented.
l The licensee's corrective action program also identified similar examples of this problem.
Some of the corrective actions for these issues were narrowly focused and lacked an
integrated effort to address the inconsistencies in application of administrative controls i
within the maintenance department.
M1.4 Troubleshooting a Breaker Indication Failure
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The inspectors reviewed the licensee's troubleshooting and corrective actions for a failure
of the control room indication for motor-driven auxiliary feedwater pump P-38A. The
associated work order Packages 9708867 and 9804735 were complete and thorough.
No administrative or technical concems were identified.
I
M1.5 Freeze Seal for Repair of Component Coolina Water Check Valve. 2CC-745 (IP 61707)
The licensee used a freeze seal to assist in the performance of a visual inspection and
repair of 2CC-745. The inspectors verified that the licensee had taken appropriate
measures to address industry-related problems with freeze seals. Maintenance
Procedure RMP 9327, "CC-745 Swing Check Vane Inspection," Revision 0, and
10 CFR 50.59 safety evaluation (SE)98-037 contained requirements that reflected these l
l measures and were determined to be adequate by the inspectors. '
l
The inspectors attended the maintenance and operations department freeze seal pre-
l~ evolution briefings held on March 17,1998. The briefings were thorough and covered
command and control responsibilities, expected communication standards, and i
'
i contingencies. Teamwork between the different disciplines was evident and participants
i displayed a good questioning attitude. The licensee completed inspection and repair of l
!
2CC-745 as planned.
l
!
M8 Miscellaneous Maintenance issues
M8.1 (Closed) LER 50-301/95006: PORV (Power-Operated Relief Valve) Post-Maintenance
l Testing Not Performed Prior to Establishing LTOP (Low Temperature Over-Pressure
l Protection). The licensee identified that LTOP was not properly established after the
i reactor vessel head was reinstalled because one of two PORVs required for LTOP was
inoperable. The valve was considered inoperable because post-maintenance testing had
not been completed. A root cause evaluation by the licensee identified that a
l
misunderstanding in the work control center resulted in the post-maintenance test for the
l valve not being performed before the reactor head was reinstalled. With the reactor head
installed, LTOP was required. In addition, control room operators were unaware that the
l
post-maintenance test had not been completed.
Two operable PORVs were required for LTOP, but the T/S allowed one valve to be
! inoperable for a limited time period. The licensee was allowed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to restore the
!
inoperable PORV and an additional 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to depressurize and vent the reactor coolant
system if the PORV could not be made operable. However, the valve was inoperable for
about 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> and the reactor coolant system had not been depressurized or vented.
Licensee management counseled operators and work control center staff on the
inappropriate delay in completing the post-maintenance test and revised several
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procedures to highlight the need for PORV operability (and establishing LTOP) prior to
reactor head installation. The problem has not reoccurred and an extensive review of the
post-maintenance testing process by the licensee in the past year, with concurrent NRC
monitoring of that review (see, for example, Sections M1.1 and M3.1 of
IR No. 50-266/97010(DRS); 50-301/97010(DRS), has given further assurance that this
will remain an isolated event. This licensee-identified and corrected violation is being
treated as a non-cited violation (NCV 50-301/98006-03(DRP)), consistent with
Section Vll.B.1 of the NRC Enforcement Policy.
M8.2 (Closed) IFl 50-266/%002-01(DRP): 50-301/96002-01(DRP): This IFl comprised several
inspector concems generally related to work planning and scheduling. One of those
concems was that an SE which addressed equipment operability could be prepared and
approved without control room personnel being made aware of possible changes to the
operability status of plant equipment specified in the SE. Because of this and other
problems with the SE process, the licensee reviewed numerous existing SEs, extensively
l restructured the goveming procedure for conducting an SE, trained plant staff on the
revised procedure, and established a multi-disciplinary team of which a member would
review all SE screenings. Recent NRC inspections (irs No. 50-266/97010(DRS);
50-301/97010(DRS) and No. 50-266/97023(DRS); 50-301/97023(DRS)) have identified
that the SE process has improved. The original concem of this IFl has been adequately
addressed.
A second concem pertained to the concurrent use of a CCW pump as the redundam
pump for two other CCW pumps. This concem was adequately addressed with the
revision (in June 1997) of T/S 15.3.3.C. for the CCW pumps. This revision removed the
previous ambiguity on redundant pumps and does not allow the use of a CCW pump
assigned to one Unit as a redundant pump for the other Unit. ,
'
The remaining two items pertained specifically to poor planning and scheduling of work
on an emergency diesel generator (EDG) and a CCW pump. Recent inspection reports
(irs No. 50-266/97003(DRP); 50-301/97003(DRP), No. 50-266/97006(DRP);
50-301/97006(DRP), No. 50-266/97013(DRP); 50-301/97013(DRP), and
No. 50-266/97021(DRP); 50-301/97021(DRP)) document additionalinstances of poor
work planning and scheduling. Although none of these items involved violations of NRC
requirements, they indicated that the work planning and scheduling process was weak
As discussed in IR No. 50-266/97006(DRP); 50-301/97006(DRP), the licensee has
recently undertaken several initiatives following an extensive maintenance program
improvement review. Because the implementation of these programmatic initiatives is
being tracked as an IFl (50-266/97006-02(DRP); 50-301/97006-02(DRP)) and the original
SE and CCW concems discussed above have been adequately addressed, the two
concerns about specific work planning and scheduling problems are considered closed.
M8.3 (Closed) LER 50-266/97042: Failure to perform containment personnel air lock
surveillance while door interlock is inoperable. The events and circumstances of this LER
were discussed in IR No. 50-266/97021(DRP); 50-301/97021(DRP), Section M2.1. A
Notice of Violation was issued regarding this matter. Therefore, this LFR is considered
l closed with the existing open violation (VIO 50-266/97021-02(DRP);
i 50-301/97021-02(DRP)) serving as the inspection tracking mechanism for completion of
the corrective actions. .
I
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lil. Enaineerina
E1 Conduct of Engineering l
l
E1.1 EDG Room Ventilation System Safety Classification
a. Inspection Scope UP 37551)
The inspectors reviewed aspects of the safety classification of the EDG room ventilation
systems.
b. Observations and Findinas
The inspectors reviewed the safety classification of the G-01 and G-02 EDG room
ventilation system components. This review was performed while independently
assessing the technical merits of an OD associated with EDG output ratings at elevated
room temperatures. The inspectors noted that the G-01 exhaust fan control panel
(C-032) was classified as safety-related; however, the G-02 exhaust fan control panel
(C-036) was classified as nonsafety-related. The inspectors questioned the system
engineer about this difference. After reviewing the component history, the system
engineer determined that the list of safety-related components had not been appropriately
updated to add C-036 as committed to in LER 50-301/91001-01. This problem was
documented in CR 98-1084.
The inspectors reviewcd the corrective actions for CR 98-1084 to ensure that the issue
had been adequately addressed. The corrective actions consisted of a broad review of
the EDG room ventilation system to determine whether any other discrepancies existed,
and a review to determine whether the condition was reportable. While these two
corrective actions were appropriate, both the CR and the corrective action documents
specified that the as-found condition was administrative in nature. The problem could
have been more substantial had the appropriate configuration and material controls not
been maintained between the time C-036 was dedicated as being safety-related and the
identification of the error. The inspectors communicated this concem to the appropriate
system engineering supervisor, who initiated an additional corrective action to review the
maintenance and modification history of C-036 to ensure that its configuration and
material status had not been compromised. No problems were identified during this
review.
1
The inspectors considered the safety significance of this specific issue to be minor;
therefore, the failure to implement effective corrective actions regarding the safety
classification of C-036 (with respect to LER 50-301/91001-01 and CR 98-1084) was a
. non-cited violation (NCV 50-301/98006-04) of 10 CFR Part 50, Appendix B, Criterion XVI,
" Corrective Action," consistent with Section IV of the NRC Enforcement Policy.
c. Conclusions
The inspectors identified a minor discrepancy in the licensee's list of safety-related
components. Specifically, a ventilation control panel in an EDG room was misclassified.
The licensee's initial corrective actions did not include determining if the ventilation
system operability had been challenged while the component was incorrectly classified as
15
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being nonsafety-related. Subsequent review indicated no problems in this area and the
discrepancy was corrected.
E1.2 Seismic Isolation in Pipina Systems
a. Inspection Scope (IP 92700)
The inspectors reviewed two LERs that dealt with discrepancies conceming the
conformance of plant piping systems to Final Safety Analysis Report (FSAR)
commitments.
b. Observations and Findinas
Licensee Event Report 50-266/97021 documented the failure to maintain two valves in
the spent fuel pool (SFP) cooling system in a normally closed position. These two valves
separated seismically qualified portions of the SFP cooling system from non-seismically
qualified portions of the system. Licensee Event Report 50-266/97028 documented that
piping which was not seismically qualified was connected to the seismically qualified
refueling water storage tank (RWST) by way of normally open valves. The FSAR requires
that valves which separate seismically qualified from non-seismically qualified systems be
normally closed. i
The licensee initiated a broad assessment of a!I systems which contained a seismically
qualified to non-qualified interface. Corrective actions were planned for all pipe systems
where adequate system separation did not exist. These actions were discussed with the j
NRC during public meetings and were documented in docketed letters to the NRC dated
July 25,1997 (NPL 97-0432), and December 19,1997 (NPL 97-0803). The inspectors l
reviewed the documentation associated with this issue and considered the docketed
information to be accurate and comprehensive. The corrective actions were considered
to be appropriate. This licensee-identified and corrected, non-repetitive failure to
maintain the SFP cooling system and RWST recirculation pipe isolation valves in the
design position (normally closed), was a non-cited violation (NCV 50-266/98006-05(DRP);
50-301/98006-05(DRP)) of 10 CFR Part 50, Appendix B, Criterion lil, " Design Control,"
cor.sistent with Section Vll.B.1 of the NRC Enforcement Policy,
c. C_o nclusions
The licensee identified and implemented effective corrective actions for two cases where
valves between seismically qualified piping systems and non-qualified piping systems
were not maintained in a closed position as required by the FSAR.
E1.3 Maintaining Desian Basis Inteority l
l
a. Inspection Scope (IP 37551)
l
l The inspectors reviewed the current status of licensee actions to ensure conformance of
plant piping systems to FSAR commitments.
16
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a
j
b. Observations and Findinas )
The inspectors discussed the status of the licensee's ongoing assessments of the
adequacy of the separation of seismically qualified and non-qualified piping systems with
the cognizant design engineering personnel. The engineers described the screening
process being used to determine whether various systems were in conformance with the
FSAR commitments, and whether modifications would be required for systems and
components which were operable, but not in compliance with the existing FSAR. The
screening criteria included the identification of motor-operated valves and check valves
which could serve the function of a normally closed valva. When such motor-operated
valves or check valves existed, additional corrective actions for those systems were not
considered necessary to address the seismically qualified to non-qualified separation
concem. However, the criteria did not require the evaluation of whether such motor-
operated valves or check valves were considered seismic-class boundsry valves in the
system design and licensing bases.
The inspectors asked whether the screening criteria had been discussed with the onsite
licensee staff performing rebaselining reviews of the inservice testing (IST) program and
the FSAR. The engineers inrifcated that such discussions had not taken place. The
I inspectors subsequently determined that the IST and inservice inspection programs could
have been affected by the seismic review program screening criteria, and that the IST ;
'
system engineer had not been aware of the seismic review until after the inspectors
questioned the design engineering personnel. While this issue may have eventually been i
identified by the licensee through supervisory reviews of the results of this seismic review
'
l program, the inspectors considered the failure to integrate the seismic review program
'
with the IST testing program review a weakness.
The licensee had several ongoing, parallel improvement initiatives which were in l
'
response to previous NRC enforcement actions. These included development of design
basis documents, a verification and update of the FSAR, rebaselining the IST program,
reviewing the inservice inspection program, rewriting system operating procedures, and
updating the IST procedures. Changes in system design basis, such as the addition of a
safety-related function to an existing valve, brought about by licensee staff working on
one of these efforts, could negatively affect the other improvement initiatives if not
properly documented and coordinated. The inspectors reviewed the licensee's response
to the latest Systematic Assessment of Licensee Performance, report
No. 50-266/97001; 50-301/97001, and found that the licensee acknowledged the need to
- . control design basis changes when making plant hardware changes, but that there was
l no specified initiative to control the effects of design basis changes that might occur as
j the result of analysis or software changes.
The inspectors met with senior licensee management to express the concem that design
engineering staff had been working on a committed corrective action for eleven months
without coordinating their efforts with other interrelated corrective action initiatives. The
inspectors asked whether this was indicative of a broad problem in design engineering, or
was an isolated incident. The licensee managers acknowledged the inspectors' concem,
and were reviewing the issue at the end of the inspection period. The inspectors will
track the licensee's actions to ensure that design basis changes, including those brought
17
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about by analysis, are property documented and communicated as an inspection follow-
up item (IFl 50 266/98006-06(DRP); 50-301/98006-06(DRP)).
c. Conclusions
Offsite, corporate office-based engineering personnel working on a corrective action
! commitment initiative to assess the adequacy of a separation of seismically qualified and
non-qualified piping systems were performing analyses and taking credit for components
l
i to function in a manner that may not have previously been considered in the design basis.
l The engineers had not evaluated whether such reliance rnight constitute a design basis
change. Additionally, onsite licensee personnel performing concurrent and interrelated
corrective action initiatives had not been informed of the potential design engineering
.
activities that could have affected the results of these other initiatives.
I
E3 Engineering Procedures and Documentation
!
E3.1 Review of Desian Basis and Controls for 125-Volt Direct Current (Vdc) System
i
a. Inspection Scope (IP 37551)
l
i
l The inspectors reviewed the design basis document and applicable battery loading
L calculations for the 125-Vdc system. The review was performed to ensure accuracy of
l the documents and verify operability of the system relative to the design basis.
l
l
b. Observations and Findinas
The 125-Vdc battery system is designed to provide service for one hour in the event of a
i total loss of altemating current voltage at the station (station blackout). During a review
l of this system, the inspectors asked licensee representatives to provide any additional
! documentation which may take into account battery loads added to the system since the
latest master calculation was generated.
l The licensee provided the inspectors with a current listing of the master, individual
l battery, and additional equipment calculations for the 125-Vdc system. Many of the
additional equipment loads were listed as evaluated but awaiting update to the goveming
battery calculations.
The inspectors held a meeting with the responsible system engineer to discuss the
system loads and calculations. The engineer stated to the inspectors that the process
used to review modification of loads against existing calculations did not prompt the
reviewer to consider other outstanding loads as the result of other modifications affecting
the system, to determine the cumulative effect on the 125-Vdc system. This condition
could have led to concurrent modifications not referencing appropriate updated battery
, loading calculations. However, the system engineer subsequently performed
l conservative calculations from the modification documentation available to illustrate that
'
the 125-Vdc system was operable. The process problems identified were described in
CR 98-1528. An OD, written on April 9,1998, indicated that the system was operable.
The inspectors regarded the lack of maintaining accurate documentation of the 125-Vdc
decign basis capabilities a violation (VIO 50-266/98006-07(DRP); 50-301/98006-07(DRP))
of 10 CFR Part 50 Appendix B, Criterion Ill, " Design Control," which requires that design
18
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\
l .
changes be subject to design control measures to assure that the design basis is
maintained. This problem dated back to about May 31,1995, when the calculations for
the Nos.105,106, and 305 station batteries had last been updated. The inspectors
reviewed the OD regarding the issues discussed above and had no further questions on )
the matter. 1
c. Conclusions l
l The inspectors concluded that tne 125-Vdc system was capable of meeting design basis ;
'
functions. However, the failure to maintain an up-to-date battery loading calculation was
considered a violation of 10 CFR Part 50, Appendix B, Criteiion ill, " Design Control."
! E3.2 Reactor Enaineerina Update of Estimated Cribcal Rate Position Calculation )
!
a. Inspection Scope (IPs 37551 and 71707)
l
l The inspectors reviewed the circumstances of the suspended critical approach on
'
March 28,1998, discussed in Section 01.1 of this report. I
b. Observations and Findinas
! As discussed in Section 01.1 of this report, the first attempt to bring the Unit 2 reactor
l critical on March 28,1998, was suspended due to the licensse's identification during the 3
withdrawal of Control Bank "D" control rods that the reactor would become critical much !
.
earlier than anticipated based on the estimated critical rate calculation. Operations I
l _ personnel followed appropriate procedures regarding this matter. The reactor critical rate )
position was recalculated prior to a second attempt to bring the reactor critical.
i
l The inspectors reviewed the information regarding the first critical approach to ascertain
l why the estimated rate calculation was in error. The inspectors leamed through
!
'
interviews of reactor engineering personnel and a review of a recent reactor engineering
self-assessment, that the estimated critical rate position calculation procedure had been
identified as needing revision. In the self-assessment report dated December 3,1997, a
finding highlighted the need for obtaining accurate xenon information from a previous
shutdown to ensure the accuracy of subsequent startup critical parameters. The reactor
engineering organization had not implemented this recommendation prior to the restart of
Unit 2. The inspectors regarded this matter as another example of a problem with reactor
l engineering performance that resulted in a previous violation
!
(VIO 50-266/98003-02(DRP); 50-301/98003-02(DRP)).
c. Conclusions
l The inspectors concluded that the reactor engineering organization provided an
inadequate critical rate position procedure to operations personnel during the startup of
Unit 2. Although reactor engineering personnel had previously identified problems with
the procedure, timely corrective actions had not been taken. The problems revealed
during the startup were considered additional examples of a previously identified problem
with reactor engineering performance, for which a violation had been recently issued.
19
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!
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'
.
l. E4 Engineering Staff Knowledge and Performance (IP 37551)
During a vertical slice review of the 125-Vdc system, the inspectors interviewed the
engineer responsible for the system. The discussion involved design bases and
operability considerations for the system. The engineer had been assigned to the system ,
for less than one month.. Nevertheless, the inspectors noted that the engineer displayed .
clear ownership of the 125-Vdc system and conveyed a sensitivity to emerging issues
affecting the system and aggressively pursued issue resolution. ;
E6 Engineering Organization and Administration
E6.1 Conduct of the Dutv Technical Advisor Proaram
a. Inspection Scope (IPs 37551 and 71707)
l
l As part of the monitoring of the Unit 2 reactor startup on March 28,1998, the inspectors
reviewed the implementation of the duty technical advisor (DTA) program.
b. Observations and Findinas
On the evening of March 27,1998, during the first attempt to t::ing the Unit 2 reactor
critical, the inspectors noted that the DTA (a reactor engineer) also served as the startup
engineer for the Unit. This individual had been the DTA for the day and was present for
, the critical approach which began around 2:00 a.m., on March 28,1998. The DTA had
! been called earlier in the evening to review and verify procedures and calculations for the )
l reactor startup. The inspectors queried the DTA as to his alertness and if he had an
opportunity for rest earlier in the evening. The DTA indicated that he was able to get
some brief rest and felt capable to oversee the reactor startup. The inspectors noted no
problems regarding the DTA's performance during the subsequent startup attempt.
The inspectors noted during the second attempt to start up Unit 2 on March 28,1998, at
around 2:00 p.m., that the same individual was serving as the DTA (but not as the startup i
l
engineer). The inspectors asked the DTA about the two consecutive days of work. The
DTA indicated that due to a reduction in the number of qualified DTAs, consecutive days
l were occasionally required.
Concemed about the appropriateness of DTAs standing 48-hour-long watches, the
inspectors reviewed the licensee's DTA program and response to NRC Generic
l
Letter 86-04, " Policy Statement on Engineering Expertise on Shift." The inspectors noted
i
that the program was approved by the NRC for DTAs to stand 24-hour watches and that
the number of available DTAs would be sufficient for adequate rotation of DTA-qualified
. personnel. The inspectors also noted that the intent of the procedure describing
l
implementation of the DTA program (Nuclear Organization Manual Duty Technical
l Advisor Procedure) was that DTAs would not serve a collateral position while functioning
l as a DTA. The inspectors acknowledged to station management that the DTA who
served as the startup engineer for the first critical attempt was not the "on-call" reactor
engineer and that this met the verbatim requirements of the procedure However, the
,
'
fact that the DTA served as the startup engineer was not in accordance with the intent of
the procedure.
20
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r 1
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. i
l
The inspectors discussed with licensee management the concems regarding DTAs
serving 48-hour shifts and their fitness-for-duty to fulfill their safety-related role in
- response to an emergency. Licensee management indicated that oversight of the ,
'
DTA program would be assigned to the operations department manager and that I
consecutive shifts would no longer be allowed. The operations manager issued an '
electronic message to all DTAs regarding this matter, following the discussion with the
inspectors.'
Licensee management also indicated that this issue would be further corrected later in
the year as plans were well underway to establish a shift technical advisor program which j
would be controlled by the operations department.
c. . Conclusions
The inspectors concluded that the practice of DTAs serving two consecutive shifts
(48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) was not consistent with the intent of program procedures and raised questions
regarding the DTA's fitness-for-duty. Although the inspectors noted no associated ;
performance issues, licensee management immediately revised expectations to preclude 1
potential fitness-for-duty issues. ;
!
E8 Miscellaneous Engineering issues
E8.1 (Closed) VIO 50-266/96002-05(DRP): 50-301/96002-05(DRP): Three examples were !
identified regarding the failure to update the FSAR as required by 10 CFR 50.71(e). The ;
licensee revised the FSAR to address the three examples and subsequently formed an '
j
interdisciplinary process improvement team to review the FSAR update process to ensure
that all required changes were being identified and implemented in a timely manner. One
outcome of the review was a revision of the FSAR change procedure (Nuclear Power
Business Unit Procedure, NP 5.2.6, "FSAR Updates"). However, during a followup
inspection of this area (IR No. 50-266/97023(DRS); 50-301/97023(DRS)), NRC inspectors
identified two additional examples where the FSAR had not been revised in a timely
manner and a violation of 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action,"
was cited. The earlier violation is considered closed and the corrective actions for the
failure of the previous long-term corrective actions will be reviewed as part of the more
recent violation (VIO 50-266/97023-03(DRS); 50-301/97023-03(DRS)).
E8.2 (Closed) VIO 50-266/96003-04(DRP): 50-301/96003-04(DRP): Contrary to American
Society of Mechanical Engineers (ASME) Code post-maintenance testing requirements,
service water pump P-32E was retumed to service in December 1995 without
determining a new vibration reference value or confirming the previous reference value.
This issue involved the retum of the pump to service with vibrations in the " alert" range.
In a letter to the NRC dated July 19,1996, the licensee did not agree that this issue was
a violation of ASME Code requirements and did not address what actions were being
taken to prevent reoccurrence of a similar problem. As discussed in a letter to the
licensee from the NRC, dated October 30,1996, the licensee has taken steps to prevent
recurrence.
Early in 1997, inspector review of the repair, testing, and retum-to-service of the P-32A
service water pump identified that the licensee still had a misunderstanding of ASME
Code reference value requirements. This misunderstanding was resolved before the
21 ,
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pump was retumed to service. Subsequent NRC review of the licensee's inservice l
testing program in mid-1997 identified no additional problems with reference value I
requirements (Section M3.1.b.4, IR No. 50-266/97010(DRS); 50-301/97010(DRS)).
E8.3 (Closed) IFl 50-266/96006-01(DRP): 50-301/96006-01(DRP): The inspectors will review
the results of the licensee's review of the inservice testing program to ensure that design
basis requirements for all safety-related pumps are incorporated in IST program test i
acceptance criteria. A followup programmatic review of this issue by NRC inspectors
(IR No. 50-266/96013(DRP); 50 301/96013(DRP)) did not identify any problems; howevel,
the inspectors kept the IFl open pending a review of the incorporation of instrument 1
- inaccuracies into IST acceptance criteria. In late 1996, the licensee completed )
I
engineering calculations addressing the incorporation of instrument inaccuracies into the l
acceptance criteria. The inspectors reviewed Calculation No. 96-0233 for the '
containment spray pumps and verified that the instrument inaccuracies had been 1
incorporated into the pump IST acceptance criteria. :
Partly because of the concems identified in the past two years by the licensee and the i
NRC, the licensee initiated an extensive rebaselining of the IST program in mid-1997. 1
The rebaselining effort was being conducted by a team of two full-time contractors, one
i
part-time contractor, and the site IST program coordinator. A brief description of the
rebaselining project was provided to the NRC in a letter dated December 12,1997, from
the licensee. To date, the project has resulted in an extensive rewriting and amplification
of IST background documents and the generation of numerous condition reports. l
E8.4 (Closed) LER 50-266/96016: Pressurizer Safety Valve Lift Setpoint Out of Tolerance Due
l
'
to Temperature Effects. This item was discussed and dispositioned in Sections E8.2
and E8.3 of IR No. 50-266/98003(DRP); 50-301/98003(DRP), but the applicable LER was
misidentified as LER 50-266/96014, which had previously been closed. This section
corrects the administrative error (referencing the incorrect LER number) contained in
IR No. 50-266/98003(DRP); 50-301/98003(DRP).
E8.5 LQlosed) LER 50-266/97021: SFP Cooling System Not in Accordance With Plant Design
,
Basis. This item is discussed and dispositioned in Section E1.2 of this report.
l
l
E8.6 (Closed) LER 50-266/97028: RWST Recirculation Piping not in Compliance with Plant
Design Basis. This item is discussed and dispositioned in Section E1.2 of this report.
l
IV. Plant Support
R1 Radiological Protection and Chemistry (RP&C) Controls
'
R1.1 Unit 1 Refuelino Outaae Radiological Controls Performance Durina This inspection Period
(IP 71750)
The licensee had recorded 85 person-rem for the Unit 1 refueling outage at the end of the
inspection period. This was on track with established goals which projected the outage
personnel exposures to total about 130 person-rem. Personnel contamination events
(PCEs) were much higher than anticipated with 81 recorded at the end of the inspection
period. The goal for the entire outage was set at 63 PCEs. Most of the PCEs were low-
l
1
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level contaminations (shoes); however, the health physics department was in the process
of evaluating the potential causes for the higher than expected number of PCEs. The
inspectors concluded that the licensee was maintaining good radiological controls for the
Unit i refueling outage and an appropriate response was being undertaken to address
higher than anticipated PCEs.
R7 Quality Assurance in RP&C Activities
R7.1 Quality Assurance Audit of Health Physics Exit Meetina (IP 7175,0.]
The inspectors attended a quality assurance department audit exit meeting on April 6,
1998. The audit included a review of various aspects of the radiation protection program
including, instrumentation controls, offsite dose calculation manual adequacies, radiation
protection personnel training, and implementation of personnel dosimeter programs. The
auditors identified several findings within these areas which were both administrative and
performance-based. The radiation protection program manager openly discussed the
findings with the auditors to gain a clear understanding of the issues. The results of this
audit will be contained in audit report A-P-98-03 which was not issued at the end of the
inspection period.
V. Manaaement Meetinas
X1 Exit Meeting Summary
The inspectors presented the inspection results to members of licensee management at the
conclusion of the inspection on April 17,1998. 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.
X3 Meeting With Local Public Officials
The inspectors, along with the Senior Resident inspector from the Kewaunee Nuclear Power
Plant, met with local officials from the Town of Two Creeks, Kewaunee County, and Manitowoc
County on Thursday April 16,1998, at the Two Creeks Town Hall in Two Creeks, Wisconsin.
The inspectors provided the officials with an overview of NRC organizations, the resident
inspector program, and the inspection process. Local officials asked the inspectors questions
i regarding these matters and other aspects of the NRC, which were answered by the inspectors.
i The officials thanked the inspectors for the opportunity to meet and ask questions.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee
Wisconsin Electric Power Company M/EPCO)
S. A. Patuiski, Site Vice President
A. J. Cavia, Plant Manager (outgoing)
M. E. Reddemann, Plant Manager (incoming)
R. G. Mende, Operations Manager
W. B. Fromm, Maintenance Manager
J. G. Schweitzer, Site Engineering Manager
R. P. Farrell, Health Physics Manager
D. F. Johnson, Regulatory Services and Licensing Manager
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INSPECTION PROCEDURES USED
IP 37551: Onsite Engineering
IP 40500: Effectiveness of Licensee Controls in identifying, Resolving, and Preventing
Problems
IP 61726: Surveillance Observations
IP 62707: Maintenance Observations
IP 71707: Plant Operations
IP 71750: Plant Support Activities
IP 92700: Onsite Follow up of Written Reports of Noaroutine Events at Power Reactor
Facilities
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
50-266/98006-01(DRP) VIO Failure to follow the procedure regarding reactor
operator observations of the main control panels
50-266/98006-02(DRP) VIO Failure to follow the procedure regarding the
weighing of the reactor vessel intemals lifting rig
50-301/98006-03(DRP) NCV Failure to perform post maintenance testing prior to
placing LTOP in service
50-301/98006-04(DRP) NCV Failure tu implement corrective action regarding C-
036
50-266/98006-05(DRP) NCV Design control of seismically controlled piping
50-301/98006-05(DRP) systems related to SFP and RWST
50-266/98006-06(DRP) IFl Followup of design basis changes to ensure
50-301/98006-06(DRP) proper documentation and interdepartmental
I communications
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50-266/98006-07(DRP) VIO Failure to implement adequate design control
50-301/98006-07(DRP) measures for 125-Vdc system calculations
i Closed
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50-266/98004 LER Reactor coolant pump lube oil collection system
50-301/98004 design nonconformance with Appendix R
Section 111.0
50-301/98002 LER Reactor coolant pump component cooling water l
retum line check valve seriously degraded 1
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50-301/95006 LER PORV post-maintenance testing not performed prior
to establishing LTOP
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50-301/98006-02(DRP) NCV Failure to perform post maintenance testing prior to i
placing LTOP in service
50-266/96002-01(DRP) IFl Scheduling and planning of work
50-301/96002-01(DRP)
50-266/97042 LER Failure to perform containment personnel air lock
surveillance
( 50-301/98006-03(DRP) NCV Failure to implement corrective action regarding C-
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036
50-266/98006-04(DRP) NCV Design control of seismically controlled piping
50-301/98006-04(DRP) systems related to SFP and RWST
l 50-266/96002-05(DRP) VIO Licensees program weakness to update FSAR
'
50-301/96002-05(DRP)
50-266/%003-04(DRP) VIO IST Weakness - ASME Code I
50-301/96003-04(DRP)
l 50-301/96006-01(DRP)
50-266/% 016 LER Pressurizer safety valve lift set point out of tolerance
due to temperature effects
50-266/97021 LER Spent fuel pool cooling system not in accordance
j with plant design basis
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l 50-266/97028 LER Refueling water storage tank recirculation piping not
l in compliance plant design basis
Discussed
50-266/97020-02(DRP) IFl Evaluate procedure upgrade program
l 50-301/97020-02(DRP)
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l 50-266/97006-02(DRP) IFl Review maintenance program improvements
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50-301/97066-02(DRP)
50-266/97021-02(DRP) VIO Failure to test containment door interlock
50-301/97021-02(DRP)
l 50-266/97023-03(DRS) VIO Failure of corrective actions for FSAR updates
50-301/97023-03(DRS)
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LIST OF ACRONYMS USED IN POINT BEACH REPORTS
AC Altemating Current
ASME American Society of Mechanical Engineers I
CCW Component Cooling Water
CFR Code of Federal Regulations
CLB Current Licensing Basis
CO Control Operator
CR Condition Report
DOS Duty Operating Supervisor j
DRP Division of Reactor Projects
DTA Duty Technical Advisors
ECCS Emergency Core Cooling System
,
EDG Emergency Diesel Generator
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ESF Engineered Safety Feature
EP Emergency Planning
FSAR- Final Safety Analysis Report
IFl inspection Follow-up Item
IP Inspection Procedure
IPE Individual Plant Evaluation
l IR inspection Report
lLRT Integrated Leak Rate Test
IST Inservice Testing
IT '7 service Test Procedure
LCO Limiting Condition for Operation
LER Licensee Event Report
LTOP Low Temperature Over-Pressure Protection
NCV Non-Cited Violation
NDE Non-Destructive Examination l
NP Nuclear Power Business Unit Procedures j
'
NRC Nuclear Regulatory Commission
01 Operating Instruction
OM Operations Manual j
OOS Out-of-Service i
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OP Operating Procedure
ORT Operations Refueling Test
PASS Post-accident Sampling System
PCE Personnel Contamination Event
POD Prompt Operability Determination '
PORV Power-Operated Relief Valve
QA Quality Assurance
RCP Reactor Coolant Pump
RMP Routine Maintenance Procedure
RP Radiation Protection
RV Reactor Vessel
RWST Refueling Water Storage Tank
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SE Safety Evaluation !
SER Safety Evaluation Report :
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SFP Spent Fuel Pool '
TDAFW Turbine Driven Auxiliary Feedwater
TS Technical Specification )
T/S Technical Specification Test
URI Unresolved item
Vdc Volt Direct Current
VIO Violation
VNCR Control Room Ventilation
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S. Patuiski -2-
i
The violations identified above are cited in the enclosed Notice of Violation (Notice), and the j
circumstances surrounding the violations are described in detailin M e enclosed report. Please
note that you are required to respond to this letter and 5:iould follow the instructions specified in
the enclosed Notice when preparing your response. Tns NRC will use your response, in part, to
determine whether further enforcement action is necesnary to ensure compliance with regulatory
requirements.
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practic'.," a copy of this letter, its
enclosures, and your response will be placed in '.he NRC Public Document Room.
Sir cerely,
/s/ Marc L. Dapas for
Geoffrey E. Grant, Director
Division of Reactor Projects
Docket Nos.: 50-266, 50-301
l
Enclosures: 1. Notice of Violation
2. Inspection Report
No. 50-266/98006(DRP);
50-301/98006(DRP)
See Attached Distribution
DOCUMENT NAME: G:\poin\ poi 98006.drp
To receive a copy of thle document, Indicate in the b3x "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure
- N* = No copy
OFFICE Rlli (;- Rlli (, Rlll ,
NAME Kunowski:dp /fAL JMcBp)pfg Grant /// k
DATE GM98 N/d98 04H95 05/05//P
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l OFFICIAL RECORD COPY
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S. Patuiski -3-
cc w/encis: R. R. Grigg, President and Chief
Operating Officer, WEPCO
l A. J. Cayia,' Plant Manager
- B. D. Burks, P.E., Director
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- Bureau of Field Operations
Cheryl L. Parrino, Chairman
Wisconsin Public Service
Commission
State Liaison Officer
Distribution:
CAC (E-Mail)
Project Mgr., NRR w/ encl
A. Beati w/ encl
J. Caldwell w/ encl
B. Clayton w/ encl
SRI Point Beach w/enci )
DRP w/enci )
TSS w/enct
DRS (2) w/encI .
Rill PRR w/enci I
l' PUBLIC IE-01 wienc!
Docket File w/enci -
GREENS
LEO (E-Mail)
DOCDESK (E-Mail)
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