IR 05000155/1994007
| ML20029E843 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/12/1994 |
| From: | Phillips M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20029E837 | List: |
| References | |
| 50-155-94-07, 50-155-94-7, NUDOCS 9405230121 | |
| Download: ML20029E843 (12) | |
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U,S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No.
50-155/94007(DRP)
Docket No.
50-155 License No.
DPR-6 Licensee:
Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Facility Name:
Big Rock Point Nuclear Plant Inspection At:
Charlevoix, Michigan Inspection Conducted: March 17 through May 5, 1994 Inspector:
R. J. Leemon C. E. Brown S.KJnti C-
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I Approved By:
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M. P. P4ffiIips, Chief Date Reactor Projects Section 2B Inspection Summarv
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JJupection on March 17 - May_5.1994 (Recort No. 50-155/94007(DRP))
Areas Inspected: A routine, unannounced inspection by the resident inspectors and others of operational safety verification, engineered safety feature system walkdown, maintenance and surveillance, engineering, and plant support-activities.
Results:
Within the areas inspected, one violation (Section 3.a) and two non-cited violations (Section 3) were identified.
The following is a summary of the licensee's performance during this inspection period:
Plant Operations Operations personnel displayed excellent professionalism and proficiency during the multiple startups and shutdowns involved with the two outages and extensive troubleshooting activities.
However, attention-to-detail was lacking in the use of procedures in placing the IPR in service and in the unexpected ESF actuations.
Maintenance All planned work was completed for the' outages; however, for several -jobs, repeat maintenance was required.
In the case of VFW-24, this may have been prevented if post-maintenance testing had been performed after the work for the first cutage.
9405230121 940512 DR ADOCK 05000155 PDR
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Engineerino Engineers were continually involved in all outage activities and supported the extensive troubleshooting activities well.
Interdepartmental communications during the inspection period showed signs of improvement.
Plant Support ALARA efforts continued to be excellent.
The security staff was professional and efficient in performing their duties.
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DETAILS 1.
Persons Contacted
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Consumers Power Comoany
- P. Donnelly, Plant Manager
- C, Bogue, Chemistry / Health Physics Manager
- G. Boss, Systems and Project Engineering Manager M. Bourassa, Senior Licensing Technologist J. Rang, Decommissioning Project Team Leader R. Rice, Nuclear Performance and Assessment Department (NPAD) Director
- R. Scheels, Planning and Scheduling Administrator W. Trubilowicz, Operations Manager D. Turner, Maintenance Manager
- G. Withrow, Plant Safety and Licensing Director
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- M. Bielinski, Training Department Representative i
- D. Gaiser, Acting Maintenance Manager
- G. Hausler, Acting Operations Manager
- G. Cheeseman, Senior NPAD Assessor The inspectors also contacted other licensee employees including members-of the technical and engineering staffs, and the reactor and auxiliary operators, j
- Denotes those attending the exit meeting on May 5, 1994.
2.
Licensee Actions on Previously Identified Items (Closed) Unresolved Item 155/91024-03: Issues. associated with MOVs.
The MOV program procedure was being revised to incorporate several necessary changes.
A new torque and thrust calculation method was being developed to more accurately predict the as-tested valve capabilities found at the plant, with the existing torque and thrust calculations to be revised.to reflect that method when it is developed and issued.
Also, weak link data which had been received from the valve vendor was to be incorporated into the calculations.
In addition, maintenance procedures were to be revised to address the new SMB actuators which have recently been ordered to replace the old SMA actuators for the M0-7051 and 7061 valves.
The licensee has scheduled completion of these items for June 1994.
This iam is closed based on planned actions.
(Closed) Violation 155/94002-01: Failure to take prompt corrective actions concerning diesel generator overloading.. As noted in the violation, corrective actions were completed by June 30, 1993, to correct other human performance issues encountered with procedural usage.
This item is considered closed.
(Closed) Violation 155/94002-02: Failure to follow procedures for the peer inspection program by not' documenting a non-conforming condition.-
A training session was conducted on February 2d, 1994, to address the
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following: documentation of the results of quality verification inspections, proper pre-job review of work packages by the quality verifier, and administrative procedure 1.6, " Quality Verification-Program." This item is considered closed.
(Closed) Unresolved Item 155/94002-03: Measured valve closure times were less than specified to prevent water hammer for the emergency condenser.
The Division of Reactor Safety reviewed the licensee's analysis of the measured data and concluded that there was no potential for water
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hammer.
This was documented in their letter to the licensee dated May 5,1994, therefore, this item is closed.
3.
Elant Operations The licensee entered the period _in a planned maintenance shutdown.
The-plant was off line on March 2, and scheduled to be back online on March 12.
However, the outage was extended to replace solenoid operated containment isolation valves (discussed in Inspection Report 50-155/94004(DRP)) and to regain acceptable water chemistry-for startup.
The licensee started the plant up on March 21 and placed the unit on line on March 22. After startup, the unit experienced problems with the turbine initial pressure regulator (IPR).
After escalating power over several days, the licensee found that turbine output was limited by the IPR to 69 megawatts-electric (MWe), 6 MWe less than full power.
Additionally, the high pressure feedwater heater bypass valve (VFW-24)
was leaking steam at the body-to-bonnet joint, Both of these items had been worked on during the outage. After the plant manager personally observed increasing steam leakage on VFW-24, he directed an immediate forced outage for repairs on April 4.
.At the completion'of all planned work, the licensee started the unit up and placed it on line on April 17, 1994. After investigating and resolving a problem with the condenser vacuum, reactor power was increased up to the. continuous maximum released level for the rest of the period.
Planned repairs were successful and the plant was operating with very low leakage and the lowest water usage rate since 1984.
a.
Operational Safety Verification (71707)
The inspectors verified that the facility was being operated in
conformance with the license and regulatory requirements and that the licensee's management was effectively carrying out its responsibilities for safe operation of the facility..The inspectors verified proper control room staffing and coordination of plant activities, verified operator adherence to procedures and technical specifications (TS), monitored the control room for abnormalities, verified that electrical power was available, observed shift turnovers, and monitored the frequency of plant and control room. visits by station management, j
The inspectors reviewed various records, such as Caution-Tag books, switching-and tagging-order files, shift logs'and surveillances, daily orders, and maintenance work orders.
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as noted below, the inspectors determined that all observed activities were acceptable.
Control Room Decorum - The inspector observed excellent operator
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performance during three plant startups and two plant shutdowns.
There was excellent attention to detail in all areas except putting the IPR in service during the startup on March 22 (discussed below) and in the unplanned engineered safety, feature (ESF) actuation (discussed in 2.b).
The evolutions proceeded in a very controlled manner with excellent verbal communications and repeat-backs.
Both contrul room and auxiliary operators performed well.
Proper decorum was maintained throughout the evolutions, from the pre-activity reactivity briefings until steady-state plant conditions were attained.
The operators continued to maintain up-to-date logs with excellent details and chronology.
Condenser Vacuum - The licensee demonstrated an excellent and
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conservative safety attitude by delaying putting the turbine on line to investigate the cause of low condenser vacuum after starting up the reactor at 4:52 a.m. April 13, 1994.
Troubleshooting activities were very extensive, including multiple system walkdowns and valve lineup verifications.
No valves were found out of position.
The licensee brought in helium leak detection equipment and spent 3 days troubleshooting the cause for the low vacuum.
During this time, excellent team work was evident among all licensee personnel. Detailed systematic investigations were performed and recorded on status sheets and plant drawings.
The problem was traced to a leaky sight-glass gasket on a float-type drain valve for the air ejector condenser. The licensee shut
the reactor down at 6:43 p.m. on April 15 to repair both the leaking valve and an identical valve off the gland seal condenser.
Repairs also included repacking the steam supply valve to an air ejector, The licensee started up the reactor on April 16 and established a good but still fluctuating vacuum. After consultation with the site management team and evaluating the situation, the plant manager authorized placing the turbine on line and increasing steam flow. As expected, this improved the condenser vacuum.
The licensee's careful and reasoned approach to investigating and resolving the problem with condenser vacuum was an excellent example of a pro-active and conservative safety philosophy.
Response to Temperature Difference - The inspector noted good
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operator performance in noticing a temperature difference between the two reactor feed pump (RFP) lube oil sumps and outboard bearings.
There was a 5 F difference between the san.: reference points on the two pumps.
Further investigation using a pyrometer and valve manipulations revealed that both RFP lube oil cooler heat exchangers needed cleaning.
Work orders were submitted and
the coolers were successfully cleaned during the forced outagc.
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Initial Pressure Regulator (IPR) - The inspector determined that
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licensee personnel failed to fully implement procedure SOP-13,
" Turbine Generator System," Revision 139.
Although Step 6.4.1.d of the procedure specified that the IPR sensing line valve (VTG-lA) be opened or checked open before resetting the IPR, the
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licensee reset the IPR with VTG-1A closed causing the turbine
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steam admission valves to slam shut and the turbine bypass valve to open to control reactor pressure.
Reactor pressure increased to 1348 psig and reactor power increased 10 percent.
During the startup on March 22, 1994, licensed operators performed poorly in putting the turbine IPR in service; however, the control room operator performed excellently in limiting the resultant transient.
When a control-room operator reset the IPR, an immediate perturbation in reactor power and pressure occurred.
Operators took immediate action to trip the IPR and take control of the turbine via the synch-governor.
Investigation revealed that VTG-1A was jammed closed.
The operator at the turbine had attempted to verify VTG-1A open by procedure.
He couldn't move the valve but had felt the piping on both sides of the valve.
Both sides felt equally warm, so he had assumed that the valve was open.
Further investigation revealed that VTG-1A had been shut when performing IPR troubleshooting during the maintenance outage during IPR adjustments.
The safety significance of the incident
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was minor as operation of the turbine bypass valve limited the
magnitude of the power and pressure transient, even without operator actions.
The IPR was subsequently placed in service.
Technical Specification 6.8.1 requires that written procedures be established, implemented, and maintained for all structures, systems, components, and safety actions defined in the Big Rock Point Quality List.
Section 5.2 of Chapter 13 of Volume 17 of the
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P4g Rock Point Quality List requires procedures for operations'and
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maintenance activities.
Section 6.4.1.d of Operating Procedure, 50P -13, " Turbine Generator System," Revision 139, requires that the operator open or check open the IPR sensing line valve, VTG-lA.
The failure to ensure that VTG-1A was open is considered an exampie of a violation of T.S. 6.8.1 (50-155/94007-01(DRP))
b.
UJ1 planned Enoineered Safety Feature (ESF) Actuations Shifting Mode Switch to RUN - The inspectors determined that
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licensee personnel failed to fully implement procedure TR-43,
" Shutdown Margin Check," Revision 21. Although Step 3.0.d.1 of the procedure required that no scram signals be present, the operators failed to clear a scram signal for the main steam isolation valve (MSIV) having been closed before shifting the mode switch to RUN, causing an unexpected reactor protection system
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(RPS) scram with no rod movement.
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On March 20, 1994, the operators were performing a procedure that demonstrates the core remains subcritical when the highest worth control rod and an adjacent rod, at a po:.ition known to contribute
>.003 Ak/k under the most reactive conditions, are fully withdrawn.
In order to proceed with control rod withdrawal, there must be no scram signals present and rod withdrawal interlocks must be clear.
The control room operators started the test and
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were able to withdraw the first control rod (the " adjacent" rod as described above) to its intended position.
However, the operators were unable to withdraw the high-worth control rod because of an active reactor-crane-position interlock (this interlock requires that the 75 ton reactor crane not be positioned over the reactor).
The withdrawn rod was then reinserted into the core. An auxiliary operator was sent to move the crane to clear the interlock.
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control room operator then attempted to withdraw both rods and again was unsuccessful.
Again, the one withdrawn rod was reinserted.
Further investigation revealed that the reactor deck
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jib crane was not plugged in (there is also a permissive interlock associated with the jib crane position).
The licensee then decided to place the mode switch in RUN (which is discussed in the i
procedure) to bypass the permissive interlock. The operators repositioned the mode switch from the REFUEL to RUN position,
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resulting in an unexpected RPS scram.
With the mode switch in the REFUEL position, the main steam isolation valve (MSIV) closed trip signal, among others, is bypassed.
In the RUN position there are-no withdrawal interlocks and no RPS trips are bypassed, j
The operators immediately determined that the MSIV being closed was the source of the scram signal.
However, there was no rod motion as the withdrawn " adjacent" rod had been reinserted earlier.
The mode switch was repositioned from RUN to REFUEL, the safety system reset, and the interlock cleared by plugging in the jib winch and positioning it to allow rod withdrawal.
TR-43 was then successfully completed.
Technical Specification 6.8.1 requires that written procedures be established, implemented, and maintained for all structures, systems, components, and safety actions defined in the Big Rock Point Quality List.
Section 5.2 of Chapter 13 of Volume 17 of the Big Rock Point Quality List requires procedures for operations and maintenance activities. Section 3.0.d.1 of Operations Procedure TR-43, " Shutdown Margin Check," Revision 21, requires "No scram signals present to inhibit withdrawal." The failure to ensure that no scram signals were present before placing the mode switch in RUN is considered an example of a violation of T.S. 6.8.1.
This violation is considered to be an isolated case, and the licensee took immediate-corrective actions as noted above. This licensee identified Severity Level IV violation is not being cited because t_he criteria specified in Section VII.B.2 of the Enforcement Policy were met.
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Inadvertent Electric Fire Pump (EFP) Start - The inspectors
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determined that the licensee performed poorly while clearing
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caution tags after completion of valve repairs, causing an unplanned EFP start.
The plant was in cold shutdown with the
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reactor coolant system-(RCS) water level lowered to below the bottom of the steam drum for valve repairs.
Before lowering the water level, the EFP and the diesel fire pump (DFP) RDS inhibit
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switches were placed to " inhibit" to prevent automatic starting.
After the steam drum level was restored to above -17 inches (EFP and DFP automatic start setpoint), the trip relays were reset in all four RDS sensor cabinets.
However, the control room operator, clearing the caution tags, failed to reset the tripped #1 FP
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module in the RDS actuation cabinet.
When the operator _ returned the RDS inhibit switch for the EFP to " Auto", the EFP s
automatically started. The operator placed the EFP inhibit switch back to INHIBIT, then reset the actuation cabinet, then stopped the EFP, and returned the inhibit switch to AUTO.
The EFP and DFP serve as part of the engineered safeguards equipment and are run routinely during surveillances by the auxiliary operators.
The cause of the unplanned start was a control operator clearing " Caution" tags without a procedure.
The operator thought he knew all he needed to clear the tags and return the EFP and DFP inhibit switches to AUTO and did not refer to the surveillance procedure.
The safety significance of the unplanned EFP start is minor as the-
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pump is routinely run during surveillances; however, the control t
of all equipment, especially ESF actuations is important.
Technical Specification 6.8.1 requires that written procedures be established, implemented, and maintained for all structures, systems, components, and safety actions defined in the Big Rock Point Quality List.
Section 5.2 of Chapter 13 of Volume 17 of_the-Big Rock Point Quality list requires procedures for. operations and maintenance activities.
The failure to identify the need for and to use a procedure thereby preventing the unexpected EFP start is a violation of-T.S.
6.8.1.
This violation is considered to be an isolated case and the licensee took immediate corrective action in the form of operation department daily orders and operator briefings.
Management also counselled the operator involved on his responsibilities.
This licensee identified Severity Level IV violation is not being cited because the criteria :pecified in Section VII.B.2 of the Enforcement Policy were met.
c.
Plant Tours The inspectors performed tours of the plant to verify system line-ups and to ascertain that the systems were operable.
During the y
tours, housekeeping and the material condition of valves, pumps,
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supports, labeling, and major system components were assessed and items needing attention were communicated to the licensee.
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Housekeepina - Housekeeping degraded during the outages, particularly in the areas adjacent to maintenance activities.
However, the licensee took immediate steps to regain the formerly good conditions as soon as circumstances allowed.
d.
Manaaement Controls The inspectors noted the frequent presence of senior management and the continuous presence of on-call management during the outages. Management was always present for the reactivity and other infrequently performed test and evolution briefings.
One violation and one non-cited violation were identified in this area.
4.
Maintenance The inspectors observed station maintenance and surveillance activities and determined whether they were conducted in accordance with approved procedures, regulatory guides, industry codes and standards, and in.
conformance with TS.
During this review, the inspectors considered the.following: (1) were approvals obtained before initiating work, (2) were instruments calibrated, (3) were. functional tests and/or calibrations performed, (4)
were quality control records properly maintained, (5) were activities accompiished by qualified perscnnel, and (6) were results within specifications and properly reviewed with any identified deficiencies properly resolved before returning components or systems to service.
The following maintenance and surveillance activities were observed:
Intermediate feedwater heater tube leak repairs
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Number 1 recirculation pump seal replaced
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Reactor depressurization system "D" main valve replaced
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Liquid poison system flange leak repair
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Emergency condenser outlet valve (MOV-7053) repair and testing
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-Replaced nine containment isolation solenoid valves
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Reactor head vent valve (M0-N004) packing leak repair
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IPR adjustments and calibration
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Turbine balancing
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Condenser vacuum fluctuation troubleshooting
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Air ejector after condenser drain trap, clean and repair
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Gland seal condenser drain trap, clean and repair
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Emergency governor exerciser testing
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Turbine high speed stop adjustment-
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VFW-24 body-to-bonnet leak repair
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Reactor depressurization system "D" drain valves (VRDS-102D and
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103D) leak repair Turbine control elements adjustment and inspection
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Turbine moisture separator leak repairs
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Core spray heat exchanger cooling water supply valve (M0-7080)
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repair Feedwater regulating bypass valve (CV-4012) repack
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Control rod drive pressure regulator (NC-18) troubleshooting
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After the March maintenance outage, the IPR was limiting turbine power to less than rated.
Additionally, the high-pressure-feedwater heater bypass valve (VFW-24) was leaking steam at the body-to-bonnet joint.
Both of these items had been worked on during the maintenance outage.
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Other repeat maintenance items from the outage included control rod drive system pressure control valve (NC-18), reactor depressurization system (RDS)
"D" train leakage, redundant core spray motor operated valve (M0-7071) seat leakage, CRD "B" (Atkomatics) flow direction control valve, and the emergency governor exerciser.
The maintenance supervisor wrote a deviation report (D-BRP-94-031) on these repeat maintenance activities and the licensee was vigorously pursuing the root cause behind the repeat maintenance activities.
The resident staff will review the resolution of this deviation report.
High Pressure Heater Bypass Valve (VFW-24) - The licensee demonstrated overall good performance during the repairs to VFW-24.
To repair a steam leak, the licensee replaced the seal between the body-and-bonnet of VFW-24 during the maintenance outage. However, no testing of the joint was accomplished before starting the plant up in March.
As pressure increased, a steam leak developed and slowly increased until the plant manager directed the plant to be shutdown on April 4.
The licensee carefully investigated the cause for the need to re-work the valve.
Step-by-step disassembly and investigation of VFW-24 revealed no readily apparent cause for the leak developing.
Assembly of the valve was verified as correct by the manufacturer's site representative.
The root cause was determined to be that the maintenance planner had specified using an inadequate festener torque on the body-to-bonnet fasteners.
The planner had selected a "mid-range" torque from the manufacturer's recommended value range.
This had been the normal practice in selecting torque values.
However, the manufacturer's site representative said that this particular type of seal (a new application for the licensee) should be initially installed with the maximum recommended torque in order to " seat" the seal. Also, a wire attaching a manufacturer's tag may have damaged the seal.
The licensee refurbished the seal seating area on the body and the bonnet and installed the new seal using the maximum recommended faster torque value.
Successful post-maintenance testing was accomplished by pressurizing the joint using a feed water pump.
The licensee demonstrated excellent maintenance practices in job performance in that no errors were made. However, the lack of a post-maintenance test after the first seal replacement was poor, especially since this was the first time this seal had been replaced since the valve was originally installed.
Turbine Balance - The licensee completed an excellent series of turbine mainienance actions by installing final balancing weights on the turbine
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rotor during the uaintenance outage.
The 1993 refueling outage had previously been voluntarily extended to accomplish major turbine
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R-refurbishment.
This resulted in extended operation and contributed to a i
continuous full released power run until the start of the planned
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d maintenance outage on March 2.
Turbine balance measurements taken after startup from the refueling outage indicated the need for minor turbine rotor balancing.
The weights were installed during the maintenance outage (the first availability since refurbishment) and the balance was excellent with vibration readings less than 3.5 mils.
No violations or deviations were identified.
5.
Engineering (37700)
The inspectors evaluated the extent to which engineering principles and evaluations were integrated into daily plant activities.
This was accomplished by assessing the technical staff's involvement in non-routine events, outage related activities, and assigned TS surveillances; by observing on-going maintenance work and troubleshooting; and by reviewing deviation investigations and root cause determinations.
The inspectors observed good overall performance by the engineering department with excellent involvement in the outage and troubleshooting
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activities.
The on-call technical assistant (OTA) was present for all plant evolutions and provided on-the-spot assistance to maintenance and operations personnel.
Additionally, good teamwork and better inter-departmental communications were evident.
No violations or deviations were identified.
6.
Plant Support a.
Chemistry - Following the outage, the licensee experienced difficulties in controlling reactor water quality, delaying reactor startup evolutions.
The level of suspended solids exceeded the licensee's turbidity limits.
These limits were established to prevent plugging of the control rod drive block filters.
The licensee formed a multi-disciplinary water quality team to solve the problem.
The team investigated numerous approaches and was instrumental in the licensee acquiring sufficient mechanical filters and transported pure water to achieve adequate cleanup.
The review of the licensee's actions to regain water quality was documented in Inspection Report 50-155/94008(DRSS).
b.
Radiation Protectio.n ALARA (as low as reasonably achievable)
.The inspector reviewed the licensee's final dose figures for the 1993 refueling outage.
The total actual exposure was 119.74 person-REM vice the initial estimate. of 155.7 person-REM.
This was the lowest total dose for a refueling outage-in the last 10 years. The low total was achieved even though the licensee
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decided to extend the refueling outage from 49 to 71 days for major turbine repairs. The ' licensee's development of an outage planning and scheduling group, months before the refueling outage, was a large factor
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in achieving this performance. Additionally, the dose reduction
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committee has continued to aggressively promoted ALARA principals. A particular example was the adoption of an employee suggestion for a new production method for changing out control rod drives (CRDs). _ Combined with the dose-reduction team's focus on pre-job activities, the suggestion resulted 'n reducing the average dose for a CRD change out from 1.94 to 0.64 REM.
The licensee continued to demonstrate excellent ALARA performance during this inspection period. All radiation work was assessed and dose i
reduction efforts applied to the work. Dose was stressed daily at the j
morning meeting and in all planning and scheduling meetings.
The estimated dose for the maintenance outage was 12.0 REM for all items
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listed on the outage planning schedule.
Estimated dose for emergent work was approximately 6.0 REM.
The total outage dose was 15.830 REM i
which demonstrated the effectiveness of the licensee's dose reduction i
efforts.
d.
Security - The licensee contract security force participated in contingency response drills on April 18-19, 1994.
No violations or deviations were identified in this ea.
7.
Management Activities j
Licensee representatives met with NRC staff in Washington, D.C.,
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on April 13, 1994. This was the third in a continuing series of meetings to discuss Big Rock Point decommissioning issues.
l The Management Safety Review Committee (MSRC) has been established
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to increase management involvement and introduce an outside perspective to nuclear safety issues at Big Rock Point and Palisades.
The group planned to meet every other month. The group will concentrate on.
operations, engineering, maintenance, support and assessment areas.
The
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focus of the next several months will be implementing the performance j
enhancement plan being developed from the diagnostic evaluation at
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Palisades.
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8.
Exit Interv_in The inspectors met with licensee representatives (denoted in paragraph 1) on May 5, 1994.
The inspectors summarized the purpose and scope of the inspection and the findings..The inspectors also discussed the likely informational content of the inspection report,.with regard to documents or processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents or processes as proprietary.
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