IR 05000266/1993014
| ML20058C949 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/17/1993 |
| From: | Tongue T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20058C945 | List: |
| References | |
| 50-266-93-14, 50-301-93-14, NUDOCS 9312020639 | |
| Download: ML20058C949 (18) | |
Text
{{#Wiki_filter:. - , . . . . k U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report Nos. 50-266/93014(DRP); 50-301/93014(DRP) ) Docket Nos. 50-266; 50-301 License No. DPR-24; DPR-27 Licensee: Wisconsin Electric Company 231 West Michigan Milwaukee, WI 53201
Facility Name: Point Beach Units 1 and 2 t Inspection At: Two Rivers, Wisconsin i Dates: September 7 through October 25, 1993
Inspectors: K. R. Jury J. Gadzala G. F. O'Dwyer T. M. Tongue Accompanying Personnel: K. K. Bristow />[
Approved B Mi - Date/
/ 1. H. longue, Actfng~ Chief i tReactor Projects Section 3A , Inspection Summary Insngction from September 7 through October 25. 1993 ' (Reports No. 50-266/93014(DRP): No. 50-301/93014(DRP) Areas inspected: Routine, unannounced inspection by resident inspectors of corrective actions on previous findings, plant operations, maintenance, , engineering, and plant support.
Results: Two violations, one non-cited violation, one unresolved item, and one inspector followup item were identified. An Executive Summary Follows.
P1 ant Operations Unit 2 was shut down for refueling outage 19 on September 25 (Paragraph 3.b).
During a leak check of a residual heat removal (RHR) valve on October 6, '
system flow was inadvertently blocked.
The reactor was defueled at the time; 9312O20639 931118 i PDR ADOCK 05000266 l G PDR
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i l '~ therefore, there was no actual loss in the ability to remove decay heat l (Paragraph 3.d).
! ! While fuel was being loaded into the reactor vessel on October 15, the air .! supply hose to the manipulator crane operator slipped off its fitting which caused the latching mechanism to fail (Paragraph 3.e).
] During performance of inservice testing on September 18, an operator failed to , notice that the differential pressure data for boric acid transfer pump 1P-4B , fell within the required action range. A condition report generated for this j issue contained a weak evaluation (Paragraph 3.f).
Maintenance
A violation was issued for a service water (SW) system isolation valve being ' inoperable for a period longer than that allowed by technical specifications ' (TS) (Paragraph 4.d).
A violation with three examples was issued for inadequate work instructions.
_ These inadequate instructions resulted in safety related equipment failures- - and a plant transient (Paragraphs 2.a, 2.d, and 4.d).
Both safety injection (SI) pumps were disabled on September 27 while
performing a leakage reduction and preventive maintenance test on the Unit 2
SI pump suction valves. Technical specifications were not clear regarding the
acceptability of disabling emergency core cooling systems when a reactor _ic i
sFutdown (Paragraph 4.c).
A non-cited violation was issued for failing to perform a post-maintenance
test on a service water isolation valve (Paragraph 4.d).
r > Enaineerina 'l
increased lead levels in the G01 emergency diesel lube oil supply resulted in j an unplanned diesel outage (Paragraph 5.a).
Two of four Unit 2 steam generator safety valves failed their lift point test ' . ' on October 11.
These failures required the removal of the remaining four safety valves for testing (Paragraph 5.b).
j , Construction continued on the new emergency diesel generator building and the ~ new diesel fuel oil system (Paragraph 5.c).
Plant Supoort ! f Major radiological performance indicators such as station dose and number of-l contaminations continued to show good trends.
However, a number of incidents-l indicative of weak radiological controls practices were identified (Paragraph - !* 6.a).
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- l During the installation of steam generator primary manways on October 14, three individuals received skin contamination which could not be readily ~ , removed (Paragraph 6.b).
Appropriate compensatory measures were taken while portions of the protected , area fence were temporarily removed to allow for laying of underground conduit l lines for the new diesel generator facility (Paragraph 6.c).
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Persons Contacted (71707) If30702)1 l !
- M. F. Baumann, Manager, Licensing and Radiological Engineering
- J. F. Becka, Regulatory Services Manager-
- J. J. Bevelacqua, Manager - Health Physics
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- F. A. Flentje, Administrative Specialist l
W. B. Fromm, Sr. Project Engineer - Plant Engineering !
- C. M. Gray, Duty Shift Superintendent j
L. D. Halverson, Site Services Manager , F. P. Hennessy, Manager - Chemistry i W. J. Herrman, Sr. Project Engineer - Construction Engineering i
- N. L. Hoefert, Manager - Production Planning
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- T. J. Koehler, Site Engineering Manager i
- T. G. Malanowski, Project Engineer, Licensing i
- G. J. Maxfield, Plant Manager
- J. A. Palmer, Manager. - Maintenance
J. C. Reisenbuechler, Manager - Operations ! J. G. Schweitzer, Maintenance Manager R. D. Seizert,-Training Manager ,
- G. R. Sherwood, Manager - Instrument & Controls
' T. G. Staskal, Sr. Project Engineer - Performance Engineerina '! Other company employees were also contacted including members of the
technical and engineering staffs, and reactor and auxiliary operators.
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- Denotes the personnel attending the management exit interview for i
summation of preliminary findings.
2.
Corrective Action on Previous Inspection Findinas and Licensee Event [ Reports (92700) (92701) (90712) { , a.
LQp.en) Unresolved Item (266/93009-02): Improper Safety Injection - (SI) Pump and Motor Oils
i This item, discussed in detail in IR 266/301/93009, involved the ! a inadvertent switching of SI-pump;1P-15A's, oil with that of its .j motor. The motor subsequently failed during its return to service { ' test.
Based on the licensee's evaluation, the failure mechanism !
was due to inadequate prelubrication of the bearings, which l resulted in immediate failure.
The oil switching and resultant l viscosity differences were determined to not be the failure-i mechanism. he root causes of the inadequate prelubrication of the bearings were identified as a combination of personnel error and inadequate instructions in the preventative maintenance task sheet. The task sheet (0045402) made no reference to proper pump motor bearing prelubrication. -Two sheets. of.the vendor manual -; were attached to the task sheet for instruction; however, a ' paragraph was crossed out that referred to the prelubrication and
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. - . - ~. -. . .- .. .= . - , . - , ! replacement of the bearings. The inadequate instructions ! contained in the task sheet is an example of a violation of 10 ! CFR 50, Appendix B, Criterion V, Instructions, Procedures, and j Drawings (301/93014-Ola). The issue of switching the oils between
the pump and motor remains unresolved pending review and i evaluation of the operability determination.
! b.
(Closed) Inspector Follow Up Item (266/92012-02): Decay Heat Removal Procedural Weaknesses j t During a review of procedures affecting decay heat removal ! capabilities, the inspector-noted several deficiencies.
Procedure i i OP-5A, " Reactor Coolant Volume Control," contained an ambiguous i step regarding the requirements for using the reactor vessel water j level standpipe for verifying level. Additionally, the control l room alarm response book contained minimal guidance on actions to i take if either the high or low reactor vessel level alarm was j received.
A weakness was also identified in the scaffold control t procedure, PBNP 3.4.16, in that, it excluded the containment from I its applicability requirements if a unit was in cold. shutdown or j refueling.
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As corrective measures, procedure OP-5A was revised to clarify the -! requirements for reactor vessel water level comparisons.
The
revised procedure directed that the independent standpipe be used l for comparison against the normal level indicators whenever a !
level change occurred. The control room alarm response book was ' also revised to provide more detailed guidance in response to high ! or low level alarms. With respect to the scaffolding issue, the i applicability statement in the scaffold control procedure was l changed to include all areas of the containment where decay heat i < removal capability could be affected. Additionally,- procedure i PBNP 3.4.8, " Transient Combustible Control," was revised to more i clearly delineate safe shutdown areas in the plant.
The inspector , reviewed the procedure revisions and had no further concerns.
! This item is closed.
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(Closed) LER 301/93-001: Failure of-Steam Generator Sample i Isolation Valve to Fully Shut
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This report describes the failure of a steam generator blowdown { sample isolation valve to fully shut on January 27, 1993.
Details , are contained in IR 266/301/93002. -The Unit 28 steam generator ~ sample line isolation valve (2MS-2084), was found to be leaking by after it was stroked during a maintenance evolution. The failure l was attributed to a minor buildup of corrosion products on the .! valve stem. When the valve was manually closed, the stroking i ' action removed the corrosion buildup.
Even though subsequent stroking evolutions yielded satisfactory results, the' valve's j ' closing spring tension was increased by about 50 lbs so that the l valve operator can overcome future corrosion product buildup.
i Subsequent testing was satisfactory. This item is closed.
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(00eni LER 266/301/93-007: Inadvertent Emergency Diesel Start and-l Loss of a Station Battery Charger l This report describes the loss of normal power to 4160 VAC ! safeguards bus lA06 and the subsequent start of the emergency diesel generator. This event was caused by personnel error and , inadequate work instructions to verify proper opening of the ! electrical disconnects or " sliders" (see IR 266/301/93013 for i details). A similar occurrence several _ days later caused voltage ! perturbation on a vital instrument bus. A Human Performance ! Enhancement System (HPES) evaluation was performed for additional i evaluation and root cause assessment of the two events.. The lack
of procedural guidance necessary to adequately perform the work is-l an example of a violation of 10 CFR 50, Appendix B, Criterion V, i " Instructions, Procedures, and Drawings"-(301/93014-Olb). A ' supplemental LER is to be issued subsequent to finalization of corrective actions.
3.
Plant Operations (71707) (71710) (93702) (60710) (40500) The inspectors evaluated licensee activities to confirm that the facility was being operated safely and in conformance with regulatory requirements.
These activities were confirmed by direct observation, facility tours, interviews ano discussions with licensee personnel and management, verification of safety system status, and review of facility records.
To verify equipment operability and compliance with TS, the inspectors reviewed shift logs, Operations' records, data sheets, instrument traces, and records of equipment malfunctions. Through work observations and discussions with Operations staff members, the inspectors verified the staff was knowledgeable of plant conditions, responded promptly and properly to alarms, adhered to procedures and applicable administrative controls, was cognizant of in progress surveillance and maintenance activities, and was aware of inoperable equipment status.
The inspectors performed channel verifications and reviewed component status and safety related parameters to verify conformance with TS.
Shift changes were observed, verifying that system status continuity was maintained and that proper. control room staffing existed.
Access to the control room was restricted and operations personnel carried out their-assigned duties in an effective manner.
Plant tours and perimeter walkdowns were conducted to verify equipment operability, assess the general condition of plant equipment, and to verify that radiological controls, fire protection controls, physical protection controls, and equipment tag out procedures were properly.
implemented.
The inspectors noted that several notable plant events were not fully recorded in the station log. Examples include the release of individuals with skin contamination and the manipulator crane failure during fuel movement. This concern was discussed with licensee
. . . management; the licensee was evaluating procedural enhancements in this area.
a.
Unit 1 Operational Status The unit continued to operate at full power during this period with the exception of power reductions for turbine valve testing and a failure of the plant process computer system.
b.
Unit 2 Operational Status The unit operated at full power until September 25 when it was shut down for refueling outage number 19.
The main generator was taken off line at 5:06 a.m. and the reactor was taken.subcritical at 6:26 a.m.
Major activities of this 36 day (scheduled) outage included a complete core off load, reactor core barrel support stand replacement, reactor coolant pump seal work, containment accident fan refurbishment, condensate and main feed pump refurbishment, main turbine refurbishment, removal of the lower containment equipment hatch, steam generator eddy current testing, and as-built walkdowns.
c.
Enaineered Safeauards Features (ESF) System Walkdown (71710) The inspectors performed a detailed walkdown of portions of the SW and SI systems in order to independently verify operability.
These system walkdowns included verification of the following items:
- Inspection of system equipment conditions.
- Ccc#irmation that the system check-off-list (COL) and operating procedures were consistent with plant drawings.
- Verification that system valves, breakers, and switches were properly aligned.
- Verification that instrumentation was properly valved in and operable.
- Verification that valves required to be locked have appropriate locking devices.
- Verification that control room switches, indications and controls were satisfactory.
- Verification that surveillance test procedures properly implement the TS surveillance requirements.
No conditions that would affect system operability were identified.
d.
Unexpected Loss of Coolant-Flow While Defueled During the valve lineup in preparation for a leak check of valve 2RH-715C on October 6, residual heat removal (RHR) system flow was inadvertently blocked. The reactor was defueled-at the time; therefore, there was no actual loss in the ability to remove decay
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The inspector was in the ! control room at the time of the event and observed the operators'
responses.
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Valve lineup verification revealed that all valves were in their ! required positions, including the minimum recirculation flow ! valves which were in service to protect the pump. Although ! shutting an isolation valve during the valve lineup was determined
to have caused the flow stoppage, operators had expected that flow should have continued through a parallel path. However, a normally closed butterfly valve in the parallel path exhibited no s leak by instead of the expected several hundred gallons per.
~ minute.
Since RHR flow was not needed, the pump was secured for'- i the duration of the test.
! i The inspector's review of the operations work plan governing this- ! evolution indicated that it did not adequately consider the affect l of the valve lineup on the operating pump. Although the inadequacy of the work plan led to isolating pump flow, the , licensee deliberately scheduled this test to be performed while ' the reactor was defueled due to its potential to affect decay heat i removal capability.
Good outage work planning, based on the , shutdown safety analysis, prevented what might otherwise have been , a significant event. The inspector discussed this event with ! plant management and had no further concerns.
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Failure of Manipulator Crane Durina Fuel Movement i While loading fuel into the reactor vessel on October 15, the f 100 psi air supply hose to the manipulator crane operator slipped l off its fitting causing the latching mechanism to fail. LA ! . previously burned fuel assembly was being lowered into the vessel when the hose became detached.
Since the latching mechanism fails j! in the latched position, loss of operating air had no effect on-l , the crane's ability to maintain the fuel' bundle latched. As the , hose detached, its take-up reel rapidly retracted causing the end
of the hose to flap freely around the reel. As a result, two hose
clamps were thrown from the hose end.
Neither clamp landed in the ! refueling cavity or the reactor core and both were retrieved.
] ! Operators finished setting the fuel bundle down and verified that l the mechanism was properly latched. The bundle was then moved to ! a staging area, away from the reactor vessel, to repair the hose
connection. The licensee determined that the hose. connection, although made up with dual hose clamps over a barbed fitting, became disconnected because the full force of the take-up reel was ! acting directly on the fittings. The design of the fittings is l primarily to restrain the hose against the internal air pressure.
! A metal webbing arrangement was rigged directly onto the hose such-l that the section of hose at the connection remained slack and did-- a not attempt to slip off the fitting. Repairs were implemented and l !
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. i ! . fuel loading was completed without further incident.
The inspector noted various supervisory. personnel, including the plant
manager, monitoring restoration activities.
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Failure of Boric Acid Transfer Pumo to Meet Acceptance Criteria' During the Unit I quarterly boric acid transfer pump and valve test (IT-17) on September 18, an operator failed to notice that i the differential pressure data for pump IP-4B fell within the l
required action range as specified in ASME Section XI code.
l During the previous Unit I refueling outage, additional pressure and flow instrumentation was installed on the boric acid piping to comply with Generic Letter 89-04 testing requirements.
Following , t the instrumentation's installation during the previous outage, - , IT-17 was performed to ensure that pump performance had not been l adversely impacted by the work. The test procedure was then r revised to incorporate the new instrumentation.
Upon procedure.
revision, the Engineering group recognized the need to establish i' new reference values during the quarterly test that'was scheduled for September 18. However, this information was not adequately t communicated to the Operations group prior to the test ! performance. A subsequent review of the test results against the l old reference values on September 21, resulted in the pump being l declared out of service even though it was actually operable.
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Since the pump was fully operable throughout this evolution, the
licensee was in full compliance with TS.
The licensee performed an additional test which established new reference values and the ! l pump was declared back in s?rvice later that day.
' > ] ~ A condition report was initiated to document this event. The _;
inspector's review of this event and the corrective actions determined that the original condition report failed to address ! the initial operator error and the lack of communications.
The l , inspectors discussed this concern with the licensee.
Licensee
management subsequently reopened the_ condition report to ensure-- ! comprehensive evaluation of the issue.
The incomplete.evalud. ion j was considered to be a corrective action program weakness.
g.
Manaaer's Supervisory Staff Meetina l ,' ! The inspector observed sessions 93-17 and 93-18 of the Manager's Supervisory Staff.
Issues discussed included proposed TS changes, reactor coolant pump seal work,.SW valve maintenance, and proposed-l voluntary entries into two limiting conditions for operation .l ' (LCOs).
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Maintenance (62703) (61726) ! I a.
Maintenance Observation l The inspectors observed safety related maintenance activities on i systems and components to ascertain that these activities were i conducted in accordance with TS, approved procedures, and l appropriate industry codes and standards. The inspectors determined that these activities did not violate LCOs and that
required redundant components were operable.
The inspectors
verified that required administrative, material, testing, and
radiological and fire prevention controls were adhered to.
i Specifically, the inspectors observed / reviewed the following !1 maintenance activities: SMP 1142, Diesel Inspection (See paragraph 5.a for details) l
N31 Source Range Nuclear Instrumentation Discriminator
. Adjustment
MI 5.1.1, (Revision 14), limitorque MOV Torque and Limit
Switch Adjustment for Gate and Globe Valves , The 2-rotor limit switches were being r eplaced in all safety
related valves with 4-rotor limit switches to improve
accuracy of valve position indication provided to operators.
SMP 1088, SI Pump P-15A Motor Connection Inspection,-Unit 2
RMP 26 (Revision 10), Reactor Coolant Pump 2P-1A
l Temporary procedure change sheets attached to this procedure used uncontrolled continuation sheets.
Although the use of ' , non-controlled continuation sheets was allowed by procedure,
the temporary change approval forms did not indicate the number of continuation sheets, nor did the continuation ! sheets have signatures or any other means to indicate i appropriate review and approval.
The inspector discussed
this weakness with plant management; additional procedural l controls will be implemented as a corrective action.
y i 2B52-421B Breaker Replacement l
A thorough and detailed work plan was used'for this l evolution. The ' inspector noted that lifted wire control ! sheets were appropriately used during the evolution.
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Surveillance Observation i The inspectors observed certain safety related surveillance activities on systems and components to ascertain that these activities were conducted in accordance with license requirements.
. For the surveillance test procedures listed below, the inspectors l determined that precautions and LCOs were adhered to, the required administrative appyovals and tag-outs were obtained prior to test ' initiation, testing was accomplished by qualified personnel in accordance with an aproved test procedure, test instrumentation
was properly calibrated, the tests were completed at the required ' frequency, and that the tests conformed to TS requirements.
Upon , test completion, the inspectors verified the recorded test data
was complete, accurate, and met TS requirements; test - ; discrepancies were properly documented and rectified; and that the systems were properly returned to service.
, Specifically, the inspectors witnessed / reviewed selected portions of the following test activittes: , ICP 4.24 (Revision 17), Nuclear Instrumentation Source Range
Channels Step 5.4.3 of the procedure specified placing the ' Storage' switch to ' Normal'. The technicians performing the evolution skipped this step because the switch was on the inside of the instrumentation drawer.
They stated that the-switch was always in the normal positiun and therefore-did-l not need repositioning.
Step 5.4.6 directed positioning the i db attenuator switch to ' Normal', tsut the switch positions
had only numerical values. Discussions with the licensee indicated that they are aware of there procedural weaknesses and were in process of correcting them.
ICP 4.lA (Revision 8), Calibration Procedure, Reactor
Protection and Safeguards Analog Racks Te'nperature
Measurement (T,, & T,), Unit 2 '
IT-525A (Revision 9), Leakage Reduction and Preventive
Maintenance Program Test of SI System, Unit 'd ! IT-515B (Revision 6), Leakage Reduction and Preventive-
Maintenance Program Test of SI Test Line and Sptay Additive l Eductor Line (Refueling), Unit 2 IT-775 (Revision 6), Spray System RWST Suction Valves
Leakage Test (Refueling), Unit 2 i .
A minor coordination issue was noted in that this. test and IT-5158 were scheduled to be performed simultaneously.
Both-required the use of a hydrostatic pump; however, only'ona )
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This test also directed measuring leak rate from a vent valve with a graduated cylinder but the procedure specified neither use of a stopwatch nor the
length of time to measure the leakage.
These weaknesses were conveyed to plant management.
l ICP 4.26 (Revision 19), Nuclear Instrumentation Power Range i
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Disablino S1 Pumos Prior to Cold Shutdown During performance of IT-525B, " Leakage Reduction and Preventive ! Maintenance Program Test of 2SI-896A&B, SI Pump Suction Valves
(Refueling), Unit 2", the duty shift supervisor noted that both ! SI pumps were procedurally required to be. locked out.
He
consequently deferred the test because he felt that it did not
provide sufficient guidance on whether a TS LC0 was applicable for
the existing plant configuration.
There were no precautions or i' limiting plant conditions in the procedure to specify the prerequisites for removing both SI pumps from service.
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Technical Specifications were not clear regarding the i acceptability of removing SI pumps from service with the reactor j subtritical. Analogous TS exist for, among others, containment , cooling and spray, component cooling water, and SW systems.
, Technical Specification 15.3.3 directs that a reactor be placed in ! a cold shutdown condition (temperature < 200 F.) if one Si pump is inoperable in excess of 72 hours.
However, there is apparently ~
no requirement for SI pumps to be operable when the unit is subcritical.
Procedure OP-3C, " Hot Shutdown to Cold Shutdown," > directs performing IT-525B, and thereby lock out both SI pumps, once primary coolant temperature has been lowered to 400 F.
A review of past shutdowns disclosed that both SI pumps were locked i out during the performance of IT-525B on several occasions with the primary system at'400 F.
! . The inspector's review of the Westinghouse loss of coolant accident safety analysis determined that there is no clear guidance on the acceptability of unrestricted unavailability of ' both SI pumps with the plant shutdown. The analysis discusses locking out SI, but only below 350 F.
It also states that SI pumps would be available with the plant shutdown above 350 F.
i However, it assumes manual 51 initiation under these conditions, with an acceptability time period of 10 minutes.
The analysis also gave credit to the availability of the SI accumulators for , mitigating the accident; however, TS treated the accumulators in
the same manner as the pumps.
Therefore, the accumulators were - not required to be available while conducting this test.
Procedure OP-3C had been revised on September 26, 1986, to add the [ performance of this leak test at 400 F.
No analysis was performed
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The ! inspector's discussions with operators indicated that a SI-pump ! would likely be restored to service within the ten minute period , discussed in the safety analysis. This issue remains unresolved !' pending the licensee's evaluation of the adequacy of the procedures and the TS (301/93014-02).
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Service Water Isolation Valve SW-LW-61 Inocerability l ! On September 10, during the performance of Inservice Test IT-72,- i " Service Water Valves (Quarterly)," an operator noted that valve ! SW-LW-61, failed to shut as required. This valve.is the inlet isolation to a non-essential service water loads piping header.
It was required to shut during an SI signal, if less than four SW pumps are operating, in order to isolate non-essential SW loads.
A maintenance work history review identified that corrective ! maintenance was' initially performed on. April 15, 1993, to replace [ valve SW-LW-61-S.
Following the maintenance, IT-72 was performed- .! and verified SW-LW-61's operability. An inspection performed by an engineer approximately one month later revealed that the
solenoid had been installed upside down. Another maintenance work , request (MWR) was subsequently initiated to correct the solenoid's . orientation.
In order to ensure adequate corrective maintenance
the second time, the engineer discussed the additional MWR's work
scope with the electrical maintenance planner that planned the ! initial work performed in April.
It was later determined that ' this MWR required only mechanics to perform the work, and the MWR i was then transferred to a mechanical maintenance planner without
communicating the solenoid's work history and scope of_the
scheduled work.
, t On June 14, maintenance was again performed to correct the { solenoid's orientation. However, the mechanics interpreted the
instructions in the MWR to mean that the instrument air l connections to the valve were reversed.
Since instructions listed in the MWR were ambiguous as to what work was actually-to be j performed, the mechanics reversed.the air connections without . considering how this would affect valve performance.
The
. inspector's review of this event identified that the. improper - maintenance was due to a lack of communications and an -! inadequately written work plan. The inadequacy of the work plan l is a violation of 10 CFR 50, Appendix B, Criterion V, . ! " Instructions, Procedures, and Drawings" (301/93014-Olc).
' When the maintenance was completed, the work package was forwarded , to Operations for review. During this review, the DSS was to [ . determine if post-maintenance testing.was necessary prior.to .; returning the valve to service.
If no testing was required, an .; explanation must be provided on the MWR. Although a' post-maintenance operability test was specified, the test was not i ,
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performed and the DSS did not provide an explanation on the MWR.
! If post-maintenance testing had been performed, the operator would i have discovered that the valve was inoperable due to the reversed i
instrument air connections prior to declaring the valve operable.
The violation for failing to perform a post-maintenance test was ! -not cited because the identification and corrective actions satisfy the criteria specified in Section VII.B of the " General Statement of Policy and Procedure for NRC Enforcement Actions," (Enforcement Policy 10 CFR Part 2, Appendix C).
, ! A subsequent inspection was performed by the engineer responsible for solenoid valves to address the adequacy of the maintenance l performed on June 14. Again, it was determined that-the solenoid remained improperly oriented. The situation was discussed with maintenance' personnel and a third MWR was initiated to correct the .; condition. On September 10, prior to this MWR being worked, IT-72-l was performed to satisfy a periodic surveillance test requirement.
j During this test SW-LW-61 failed to shut. As stated abo've, a i review of previous work showed that this valve had been unable to i perform its intended saiety function for a period'of 88 days.
-! This is a violation of TS 15.3.3.D.I.b which requires that all r necessary valves required for the functioning of the SW system
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during accident conditions be operable.
(301/93014-03(DRP)). The inoperability of this valve was determined to be of low safety f significace, in part due to a redundant flow isolation valve (SW-
LW-62) on'the outlet of the non-essential loads' common piping l header that remained fully operable. Therefore, multiple failures ! would have been required to reduce required SW flow below minimum ! . requirements.
Because the non-essential loads' piping is seismic ! class III between the two isolation valves, failure of this piping j was also evaluated with respect to its affect on the design basis . accident.
The probability of a design basis earthquake in close i succession to the design basis accident-is very low (about E-9)' l . Additionally, a piping walkdown by the licensee indicated that
these class III piping sections were sufficiently restrained to j either meet seismic code requirements or operability criteria ! (i.e. the piping would not be expected to fail).
-l -! The licensee initiated a number of corrective actions in response i to this event. The DSS involved was counselled by the Operations ! Manager.
An HPES evaluation was performed to determine the-j issue's root causes. The licensee has also committed to perform a
quality assurance audit of completed MWRs to verify the adequacy ! of work plans, pre-job reviews, and closecut reviews with regards 'i to post-maintenance testing. This audit was expected to be ! completed by February 28, 1994. Although not a-corrective action { to this event, procedure PBNP 5.17, " Standards for MWR Work $ Plans," was issued on July 30, 1993, to ensure that adequate ! guidance exists to aid planners in drafting MWR work plans.
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Enqineerina (71707) l The inspectors evaluated engineering and technical support activities to ! determine their involvement and support.of facility operations. This j was accomplished during the course of routine evaluation of facility
events and concerns, through direct observation of activities, and ! discussions with engineering personnel.
I ' a.
Hiah lead Content in G01 Emeraency Diesel Lube Oil On September 9, a sample of G01 Emergency Diesel lube oil revealed
a lead content of 113 ppm, indicating the engine was in j " distress". The lead levels had been trending upward since G01's i overhaul in February 1993. The magnitude of the 113 ppm reading ! was compared to historical data and to normal industry averages i and determined to be abnormal.
Based upon an analysis of all i trended engine parameters, the high. lead content alone did not } necessitate declaring G01 inoperable.
However, since the reading j was.above the licensee's lead level limit and vendor's action j levels, G01 was removed from service and the engine bearings were l inspected for possible lead sources.
_l ! During this inspection, the #20 piston rod basket was discovered
to be broken. The licensee believes that the broken basket acted
on the backside of the #20 upper connecting rod bearing and as a
result, was the main contributor to the increased lead levels.
l However, the degree to which this may have contributed to the i elevated lead levels could not be determined.
Extensive , engineering support was evident throughout this evolution. Major
work performed included: replacing all the connecting rod bearings even though none were in-a disqualified condition (three i showed signs of distress); replacing one upper main bearing and
all lower main bearings; replacing the lef t side camshaft ! ' bearings; the #20 " power package" consisting of the cylinder, i piston, and connecting rod was replaced due. to its broken rod e basket; and, the engine oil and all oil filters were changed.
. Some excessive bearing wear. was identified. Although most of the
wear would not disqualify the bearings for further use, it is ' believed they contributed to the high lead content. The_ bearings -{ were replaced due to their age and to prevent further lead- -i degradation and additional. oil contamination. The engine was j satisfactorily. tested and returned to ser' rice.
Post maintenance
lead levels were measured between 0 and 4 ppm.
! ! b.
Improoer lift Settinas on Steam Generator Safety Valves i ! On'0ctober ll, a Unit 2 steam generator safety-valve failed its j lift point test.
Two safety valves are normally sent offsite each
refueling outage to verify that they relieve at the proper set ! _ point.
Because one of these two valves failed its lift test, two .. additional valves were removed for testing. One' valve in the
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. , . second group also failed its lift test.
This necessitated l removing the remaining four safety valves' for testing, all of l which subsequently lifted at their proper setpoints.
The two valves that failed the test were 2MS-2008 and 2MS-2010.
l 2MS-2008 has a setpoint of 1125 psi and relieved 6% above it.
2MS-2010 has a setpoint of 1085 psi and relieved 4% above 'it.
The l i ASME code acceptance criteria provides for a maximum 3% range above the setpoint. The licensee was evaluating the consequences
of operation with the valves at their as-found settings and- ! possible root causes. The inspector will review this evaluation - and document the results of this followup item in a future report j (301/93014-04).
J l c.
Construction of New Emeroency Diesel Generator Buildina
Construction of the building to house two new emergency diesel-- i generators and the new diesel. fuel oil system began the week of i June 7.
Initial observations of this activity are discussed in
IR 266/301/93011. During this inspection period, concrete pours continued for the main building floor and ground level walls.
Conduit for the cable runs from the new diesel generator building u to the existing safeguards busses was laid down in trenches and i backfilled with concrete up to the turbine building entrance.
i ' The inspectors monitored excavation and grading activities, concrete placement activities including installation of steel
reinforcement bar (rebar), laying of conduit, batch plant i operation, concrete transport and pumping, testing, form removal i and post pour inspection. Discussions were held with craft - , workers and supervisors to evaluate their knowledge of the job ' requirements. The inspectors will continue to monitor progress of this construction.
_: > 6.
Plant Support j - a.
Radiolooical Practices and Exposure Control - e Point Beach continued to make good progress in reducing personnel - exposure to radiation.
Cumulative exposure for 1993 was running i about-30 Rem below the levels achieved in 1992._ This marks the
fourth consecutive year of declining exposure le_vels.
The
exposure for the current Unit 2 refueling outage during this.
inspection period was 84 Rem.
This level was on course to achieve i ! the outage goal of 110 Rem.
Of negative note, the inspectors noted a number of recent licensee- ' identified weak radiological control practices.
Exuples include: degradation of a radiation area barrier rope; two instances of ., visiting contractors not wearing proper dosimetry in the ! radiologically controlled area; several personnel receiving skin contamination while installing steam generator primary manways (as i
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i detailed later in this report); and one person being contaminated by a highly irradiated particle while removing his anti-contamination clothing. Although overall program indicators such as personnel exposure continued to exhibit improving trends, the inspectors expressed concern that the number of individual events appears to have recently increased, which could eventually lead to an overall decline in performance. The potential for a ' performance decline, based on weak radiological control practices, ' was discussed with plant management and will continue to be i evaluated by the in:pectors, b.
Personnel Contaminations ! During installation of steam generator primary manways on October 14, three personnel received skin contamination which could not be readily removed. The personnel were wearing cotton gloves while handling bolts coated with a colloidal solution of graphite. This fluid, which was contaminated from the bolts, i soaked through the cotton gloves and contaminated their fingers.
? The personnel were not wearing rubber gloves because in past refueling outages, these bolts had routinely been cleaned upon initial removal of the manways and use of cotton gloves was previously acceptable. This process change was not adequately - evaluated from a radiological standpoint to ensure that rubber , gloves were specified for the workers involved.
! After employing standard skin decontamination techniques, the
fingers of the three individuals measured between 500 and 2800 cpm ' on contact (equating to about 0.3 mrem /hr). Their fingers we.>
' subsequently coated with skin cream to further aid in removing the contaminants, gloves were donnad to prevent inadvertent removal of the cream, and the individuals were sent home. Upon their return , to work the next day, the contact reading on their fingers had , been reduced by half. Subsequent decontamination was fully successful on one individual and reduced the contact readings on j the other two to about 600 cpm.
, ! At this point, it was decided to attempt decontamination of the remaining two individuals using potassium permanganate.
Licensee , procedures require that this be done under a doctor's supervision, so both men were sent to Two Rivers Community Hospital for - treatment. This process reduced the levels to about 200 cpm and
subsequent hand washing removed the remaining contamination. No
regulatory limits were exceeded during the course of this event.
i The licensee promulgated additional guidance regarding evaluations of radiological precautions for future work.
c.
Security i The inspectors, by direct observation and interview, verified that portions of the physical security program were being implemented j in accordance with the station security plan. This included-
l t , - i
. - P . checks that identification badges were properly displayed, vital areas were locked and alarmed, and personnel and packages entering the protected area were appropriately searched.
Portions of the protected area fence were temporarily removed to allow for laying of underground conduit lines for the new diesel i generator facility. Temporary fence was erected and the area was compensated with additional security personnel. The licensee's
plans and contingency measures were communicated to Region III prior to the fencing modification. The inspectors monitored the
compensatory measures enacted, questioned individual security personnel on their assigned duties in this area, and reviewed the-i changes to the security plan. No concerns were noted.
7.
Manaaement Meetinas (30702) Meetings were held between NRC Region III management and licensee management on September 9 and 10, to discuss items of interest.
Items of discussion included matericl condition, maintenance backlog, . planning, engineering support, personnel resources, and management effectiveness.
8.
Outstandina Items (92701) Inspection Follow Up Items ' Inspection follow up items are matters which have been discussed with licensee management, will be reviewed further by the inspector, and involve some action on the part of the NRC, licensee or both. A follow up item disclosed during the inspection is discussed in paragraph 5.b.
Unresolved Items , Unresolved items are matters about which more information is required in - order to ascertain whether they are acceptable items, items of noncompliance, or deviations. An unresolved item disclosed during the inspection is discussed in paragraph 4.c.
9.
Exit Interview (71707) A verbal summary of preliminary findings was provided to the Wisconsin Electric representatives denoted in Section 1 on October 28, at the ' conclusion of the inspection.
Information. highlighted during the-meeting is contained in the Executive Summary and dissenting comments were not received.
No written inspection material was provided to , company personnel during the inspection.
l The likely informational content of-the inspection report with regard to documents or processes reviewed during the. inspection was also , discussed. Wisconsin Electric management did not-identify any documents
or processes that were reported on as proprietary.
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