IR 05000155/1987002
| ML20210U002 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/06/1987 |
| From: | Jackiw I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20210T924 | List: |
| References | |
| 50-155-87-02, 50-155-87-2, NUDOCS 8702180363 | |
| Download: ML20210U002 (23) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-155/87002(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: November 27, 1986 through January 27, 1987 Inspector: S. Guthrie Approved By: ' Ids (i ,ih A l'/7 Projects S ction 2C Date p
Inspection Summary Inspection on November 27, 1986 through January 27, 1987 (Report No. 50-155/87002(DRP))
Areas Inspected: Routine, unannounced inspection conducted by the Senior Resident Inspector of Operational Safety, Maintenance Operation, Surveillance Operation, Reactor Trips, Licensee Event Report Followup, and Licensing Action Results: Of the six areas inspected, no violations or deviation were identified. No significant safety items were identifie g2] % [5$ g5 0 ,
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DETAILS 1. Persons Contacted
- D. Hoffman, Plant Superintendent
- G. Petitjean, Planning and Administrative Services Superintendent
- G. Withrow, Engineering Maintenance Superintendent
- R. Alexander, Technical Engineer
- R.~Abel, Production and Plant Performanc Superintendent L. Monshor, Quality Assurance Superintendent R. Barnhart, Senior Quality Assurance Administrator P. Donnelly . Senior Review Supervisor, Nuclear Activities Dep j D. Staton, Shift Supervisor W. Trubilowicz, Operations Supervisor
- J. Beer, Chemistry / Health Physics ~ Superintendent E. Evans, Senior Engineer J. Tilton, General Engineer D. Kelly, Maintenance Supervisor D. Ball, Maintenance Supervisor W. Blosh, Maintenance Engineer M. Acker, Senior Engineer J. Toskey, General Engineer G. Boss, Reactor Engineer L. Darrah, Shift Supervisor J. Horan, Shift Supervisor R. May, Shift Supervisor R. Scheels, Shift Supervisor J. Warner, Property Protection Supervisor
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- T. Fisher, Senior Quality Assurance Administrator R. Krchmar, General Quality Assurance Analyst
- E. Ractborski, Planning and Scheduling Administrator l
l The inspector also contacted other licensee personnel in the Operations, 1 Maintenance, Radiation Protection and Technical Department * Denotes those present at exit intervie . Operational Safety Verification The inspector observed control room operations, reviewed applicable logs i
and conducted discussions with control room operators during the inspection period. The inspector ~ verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the containment sphere and turbine building were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that
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maintenance requests had been initiated for equipment in need of- '
maintenance. .The inspector, by observation and' direct interview, i verified that the physical security plan was being implemented in ;
accordance with the station security pla '
j The inspector observed plant housekeeping / cleanliness conditions and verified implementation of radiation protection controls.- During the inspection period, the inspector walked down the accessible portions
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of the Liquid Poison,-Emergency Condenser, Reactor Depressurization, 7 Post Incident, Core Spray and Containment Spray systems to verify operability. The inspector also' witnessed portions of.the radioactive ,
waste system controls associated with radwaste shipments and barre 11n On December 3, the inspector participated with the licensee, NRR, and Region III Management in-the third conference call on the subject of '
fuel inspections for leaks during the 1987 refueling outage.. Details describing the licensee's decision not to perform fuel . inspections- '
of fuel known to be leaking, the -inspector's concerns, and the content of the two previous conference calls are presented in Inspection . ;
Report No. 155/86014(DRP), Section 3.f In the December 3 call,.the i licensee defended their original decision to_ forego inspections. .. The '
licensee did commit to have the fuel vendor conduct ultrasonic testing !
of two test bundles, one known to be leak free and a second known to [
have leaks, in a demonstration of the ultrasonic technique. These two ;
bundles would be in addition to two bundles being tested by the fuel l vendor for their own research. The licensee committed to evaluate !
the results of the demonstration and, depending on the accuracy of ;
the technique, evaluate the option.of examining up to twenty bundle At the time of the call, the licensee was not prepared to describe the criteria on which the inspection decision would be based. The .i licensee did commit, that, if additional bundles are inspected using ultrasound, the bundles would be among those being returned to the core and would be selected on a priority basis based on likelihood of leakag During the period January 12-17, the inspector observed the activities- l of the licensee's ful vendor while performing ultrasonic testing of i fuel bundles within the spent fuel pool.. The. vendor-examined four I fuel bundles, the leakage characteristics of which were known to the- -
Itcensee but not to the testing crew, in a successful attempt-to demonstrate the validity of the testing methodology. Results of ultrasonic testing of each fuel bundle was consistent with data- !
obtained from fuel sipping tests for that bundl .
During the final two days of vendor testing, the licensee examined 21 additional fuel bundles scheduled for return to the reactor'for use :
during the upcoming cycle. -The tests were conducted in response to- !
inspector concerns over the licensee's decision to forego fuel sipping ]
during the refueling outage although minor fuel failures during th '
last three cycles indicated the possibility existed that defective fuel would be returned to the reactor for further exposure. The
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inspector's concerns are presented in Section 3.f of Inspection Report No. 155/86014(DRP). The 21 bundles examined were selected among those highly suspect "H" series bundles with the greatest exposure in the core. No leaking bundles were identifie b. At the close of the previous inspection period, the licensee controlled a packing leak on Valve VP-13, the liquid poison to reactor vessel isolation valve, by placing the valve on its backseat. Details of those activities were presented in Section 3.1 of Inspection Report No. 155/86014(DRP). During the present inspection period the inspector questioned the licensee on the consistency of plant operators when placing valves in the fully open position and why VP-13 was not originally placed on its backseat when last opene The licensee informed the inspector that, in years past, valve lineup sheets included identification of those valves equipped with backseats to assist operators in using the backseat when available. However, in recent years an accurate listing of valves equipped with backseats has become outdated. The licensee recognizes the added margin of valve leakage integrity provided by the backseat compared to valves which rely only on stem packing to provide the pressure boundary and has initiated a program to research valves installed in the plant and generate an accurate list of valves with backseats. That knowledge will be factored into instructions to plant operators on valve positioning. A licensee review of operator practices when placing valves in the fully open position indicates that presently most operators open a valve fully then close the valve approximately one-half times to prevent thermal binding of the valve, c. During the period December 12-22, the plant experienced gradually increasing values for unidentified leakage from the reactor water inventory. Leak rate calculations, which are performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, increased to approximately .62 gpm, with one calculation on December 21 as high as .70 gpm. Technical Specification Require the unit to be shutdown when unidentified leakage exceeds 1.0 gpm and Big Rock Administrative Procedures require power reduction and investigation at 0.8 gpm. In response to the observed increasing trend, the licensee on December 22, reduced reactor power from the all rods all out maximum of 49 MWe to approximately 5 MWe to permit entry into the recirculation pump room and steam drum enclosure for leak investigation. A survey identified several minor packing leaks on small instrument isolation valves and a significant packing leak on VN-305, a ten inch spring assisted swing check valve in feedwater piping to the steam drum. Leakage was estimated at one liter / minute, a volume consistent with observed increases in unidentified leak rate calculation During the power reduction for inspection and repairs to VN-305, the inspector observed control room operators and maintenance personnel in the plant. Operators inserted control rods in accordance with applicable procedures, using working copies of approved rod movement sequences and laminated cards used to track individual rod notch
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movement Both licensed reactor operators communicated effectively to verify each rod manipulation. Actions of operators observed during the power reduction displayed familiarity with procedural requirements and the operational characteristics of individual plant components as evidenced by responding decisively to fluctuations in feedwater flow affecting steam drum level and appropriate setting of the turbine's initial pressure regulator and synchronous governor control for plant pressure control. The inspector observed plant maintenance and radiation protection personnel responding to the VFW-305 lea Efforts to tighten valve packing were unsuccessful and a prefabricated drip collection pan was mounted beneath the valve. Flow was directed out of the recirculating pump room to a floor drain accessible to operators for daily calculation of leak rate. Given the ability to identify and quantify leakage from the valve, the calculated leak rate was no longer included in the unidentified category. Job preplanning was evident in the equipment prefabrication and pre-stagin Radiation Protection Technician involvement was evident through the direct involvement of the technicians in supervising the entry into an area normally off limits due to excessive radiation levels and the monitoring of doses for personnel involve Initial calculations for leakage through the collection pan indicated approximately .07 gpm, a figure consistent with the overall calculated unidentified leak rate reduction from .620 gpm to .555 gpm following installation of the collection device. The unit was restored to service December 23, by withdrawing control rods to the all rods all out configuratio d. During the week of December 29, the licensee made temporary changes to Procedure 0-FFI-1, Fuel Bundle Removal, which would permit offloading of fuel from the reactor during the refueling outage with minimum nuclear instrumentation consisting of one startup range channel and one power range channel. These minimum requirements are consistent with Technical Specification 7.3.5, Extended Shutdown, but are inadequate to meet the requirements of Technical Specification 7.4.(c),
Refueling Operations, which requires "both startup nuclear instrumentation shall be in service . . . during all refueling operations." Technical Specification defines " Refueling Operation" as "any operation. . . during which a core alteration, or other operation which might increase core reactivity, is in progress."
The licensee, based their decision to unload the core with one startup range and one power range channel on their interpretation that unloading fuel, did not constitute a refueling operation because there was virtually no itkalthood of causing a net increase in core reactivity by decreasing the amount of fuel in the reacto On January 5, and before fuel offload commenced the inspector, with the concurrence of Region III management and NRR, informed the licensee that the decision to defuel with less than two startup range channels did not meet the requirements of Technical Specification The licensee was informed that based on the definition of Refueling
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Operation as a core alteration in progress, fuel offload activity represented a refueling operation for which two startup range channels are required. The Reactor Engineering Branch of NRR noted that core physics models exist which would refute the generalization that fuel offload activity could never result in an increase in core reactivity, and that an accident involving a dropped fuel bundle would increase reactivity in the core. The licensee revised procedure 0-FFI-1 to require the operability of both startup channels for all fuel offloading activity. The revision included a provision for use of fully operable and calibrated " dunker" type detectors when installed
! startup range instrumentation is not operable, an option that meets NRR approval. Unable to verify the operability of Channel 6, the licensee installed a dunker type detector and verified the operability of startup Channels 7 and The licensee began removing fuel from the vessel in the early hours of January 6, but halted movement when the fuel transfer cask required adjustment. Fuel movement resumed several hours later and was completed on January e. On January 2, the inspector observed the shutdown of the reactor to enter the 1987 refueling outage. Control rod insertion was performed l in accordance with procedural requirements and utilized the laminated i card system and approved sequences to track rod notch insertions, l
The turbine stop valve, which had displayed a recent history of failure to operate from the control room, operated normally. The
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Reactor Engineer was present in the control room. Following l 1 solation of the turbine and with startup range instrumentation
- inserted in the reactor, operators paused in the rod insertion I sequence to diagnose and correct discrepancies in readings between l Channels 6 and 7 in the startup range instrumentation and Channel 2 I
compared to Channels 1 and 3 in the power range instrumentation, i
Both Channels 2 and 6 had displayed abnormal behavior earlier in the operating cycle, with erratic behavior of channel two causing a scram on July 1, 1986. While Technical Specifications have no requirements for startup range instrumentation during power operation, Section 6.1.5.h permits shutdown operation with only one of the two startup range channels. Technical Specification 6.1.5.e permits any one of the three power range channels to be removed from service for maintenance during reactor operations. The licensee determined that i prior to resumption of rod insertion a trip signal would be inserted l in Channel 2 and repairs of Channel 6 would commenc While inserting the trip signal into power range Channel 2, the reactor scrammed on upscale /downscale readings between power range
- channels. An upscale /downscale trip occurs when one instrument is ( tripped low at 5% while another is tripped high at 120% reactor power.
l Because the plant's event recorders were not designed to identify the
! downscale signals, operators were unable to accurately determine which channel's erratic behavior had caused the trip signal. At low power l levels, all channels of nuclear instrumentation are extremely sensitive l
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to electrical noise and activity in one channel has in the past caused sympathetic actuations of adjacent channels. Following the scram, all rods inserted, the remainder of the Reactor Protection System (RPS)
functioned normally, and no engineered safety features were actuate Operator action was appropriate. Shortly after the scram, a second RPS actuation occurred, resulting in no rod motion. The signal had its origin in the Intermediate Range Channel 5 and operators concluded that, like the original RPS actuation, the cause was spurious and erratic behavior related to the instrumentation's sensitivity to electrical noise at low power levels. The licensee made the appropriate notifications to NRC headquarters, On January 10, the inspector reviewed the licensee's proposed procedural changes to Standard Operating Procedures (SOP) 2, Refueling Operation, and 50P 44, Spent Fuel Pool Operation and
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New Fuel Handling. Each procedure contains a requirement that l " containment integrity shall be in effect when moving fuel."
l Technical Specification 3.6, Containment Requirements, requires
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sphere integrity "during power operation, refueling operation, shutdown and cold shutdown conditions except as specified by a i system of procedures and controls to be established for occasions l containment must be breached during cold shutdown." The licensee's concern focused on the restriction in 50P 44 which prevented " fuel moves" without containment integrity intact. A conservative interpretation would imply no moves of fuel within the spent fuel pool to support the underwater testing activities of the fuel vendor without containment integrity. Although sphere integrity was at the
, time intact, scheduled local leak rate testing on drain isolation I
valves could result in a technical deficiency in sphere integrity if a valve should fail. Technical Specification 4.2.11(b) specifically permits fuel assemblies to "be placed in inspection stations located l in the spent fuel storage pool for inspection, detection of failed i fuel, exchange of fuel pins, or similar purposes."
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The licensee determined that the requirements of S0P 44 were overly restrictive beyond the original intent of the procedure. Both SOPS 2 and 44 were revised to remove the vague term " moving fuel" and replaced it with a requirement that " containment integrity shall be in effect when in ' Refueling Operation" as referenced in Tech l '
Spec 1.2.3." The Refueling Operation definition addressed core alterations and other activities within the reactor vessel and does not reference the spent fuel pool. The licensee addressed the need l
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to restrict releases of air "from the containment atmosphere to the outside environment" during activities underway in the spent fuel pool by further amending S0P 44 to require "that the containment ventilation valve isolation system and all air locks are operable and the plant is in cold shutdown." Given those conditions, S0P 44 changes authorize fuel moves only within the spent fuel pool and casks, components ar.d equipment moves within and out of the spent
- fuel pool. In summary, the changes will permit fuel to be moved l within the pool and cask and equipment to be removed from the pool
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under sphere conditions where direct airborne release paths to atmosphere remain intact but certain pump or drain isolation valves
.nay be tested or positioned so as to reduce containment integrity.
- Full containment integrity will continue to be required for all other plant conditions and activities. The inspector concluded that the procedure changes were consistent with Technical Specification requirements and were made in accordance with administrative
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, requirements, including PRC approval.
I g. During a one week period commencing January 7, the licensee experienced incidents of individuals working on the reactor deck becoming contaminated. Whole body counts performed daily on the
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16 individuals indicated that, based on trending of the half-life of the Cobalt-60 isotope identified the contaminations were externa At the time of the increase in contamination incidents, all fuel bundles had been offluaded and operators were involved in inspections l
and movements of vessel structural components, channels and blades.
l On January 11, the licensee halted activity on the refueling deck j pending investigation of the source of contamination. Operators
! reported the amount of iron oxide in the reactor water was unusually l high, resulting in poor water clarity and adherence of radioactive I material to handling devices used in the vessel. Site health
physics personnel instituted added controls on the refueling deck, including frequent washing of tools and cables, and assigned decontamination workers to the are On January 13, the licensee modified the normal alignment of the new spent fuel pool filter system in an attempt to filter reactor water to improve clarity and reduce activity levels. The filtration system, l
which is a new installation never before tested or operated in the facility, normally takes a suction in and discharges to the spent fuel pool via a ten micron filter in a canister. The canister is l not watertight, since any leakage through the lid would normally be
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the same spent fuel pool water being drawn up the suction hose. In i its altered configuration for use in the reactor vessel, which is physically separated from the spent fuel pool, the filter and pump stayed submerged in the fuel pool and suction and discharge hoses to the reactor vessel were laid across the reactor deck. During the approximate five minute period of operation leakage of spent fuel pool water into the water being pumped from and returned to the reactor at 250 gpm caused the spent fuel pool level to visibly decrease by several inches, and the reactor vessel shield tank to overflow via the installed standpipe to the area beneath the vessel as water was transferred from the spent fuel to the reactor vessel.
, The inspector expressed a concern that the application of a new
! system, the operating characteristics of which are unknown to
! operators and the reliability of which has not been demonstrated, offer the risk of producing unanticipated results that place the integrity of the plant in jeopardy. In this instance, the filtration system was in an untested configuration without benefit of procedural i l
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guidance and displayed leak' age characteristics in excess of those advertised by the vendor. Had alert operators not noticed falling J fuel pool level and recognized the high volume siphon effect of the hose arrangement, a significant amount of water covering spent fuel 1 could have been transferred to the reactor vessel with serious safety implication h. On January 13, the inspector discussed with licensee security personnel the procedure for devitalization of the pipe tunnel during the refueling outage to permit disassembly of components which, when disassembled, permit access to the pipe tunne The devitalization was conducted in accordance with the security pla i. On January 6, the licensee security force responded to observed malfunctions in area access control devices. The licensee responded in accordance with the security plan and supplemented the security l staff to ensure vital area integrit J. During the inspection period, the inspector reviewed with the licensee their decision to remove the lock from the high radiation area barrier surrounding the turbine. The licensee based their decision on ALARA
, considerations, noting the time necessary to remove the lock from the l barrier during which the operator is exposed to a radiation field of
[ approximately 30 mr/hr at full power. Once within the barrier the
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majority of the operator's activities are conducted at the end of the turbine housing the turbine control mechanism in a radiation field of approximately 10 mr/hr. The only location within the
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boundaried area exceeding the 100 mr/hr criteria for de<ignation l as a high radiation area is a portion of the moisture separator situated atop the turbines and not readily accessible from floor
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l level. Operators, during their normal plant tours, are required i to enter the boundaried area once each eight hour shift to perform equipment checks.
l The inspector discussed with the licensee the requirements of 10 CFR 20.203(c)(2) which requires that the entrance to a high radiation area shall be controlled, alarmed, or locked. The inspector concluded that Technical Specification 6.12, High Radiation Area, Amendment No. 63 dated January 4, 1984, relieves the licensee of compliance with 10 CFR 20.203(c)(2) by permitting,
"in lieu of the ' control device' or ' alarm signal,"' an unlocked barricade with conspicuous posting as a high radiation area and entrance controlled by Radiation Work Permit for radiation levels l
less than 1000 mr/hr at 18 inches. Locked doors for high radiation l area control are not required until radiation levels at 18 inches i exceed 1000 mr/h k. During the inspection period, the inspector reviewed the licensee's use of butt splice connectors manufactured by the AMP company after I those connectors from another licensee failed lab tests by exhibiting excessive current leakage when exposed to a steam environment. A 9 ,
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review of the licensee's electrical pen'etration scheme indicated i . extensive use of AMP splices. The licensee has recently implemented
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a program utilizing Ray Chem heat sh'rinkable tubing as insulation i for butt splices for all future applications. The inspector learned i that 382 splices inside and outside containment using connectors-i manufactured by AMP and other suppliers had been wrapped in tape ,
manufactured by 3M for which the licensee' has on file a qualification
! report. The taping was performed in 1978 in response to licensee initiated concerns over the ability of splices in ,the inner and outer cable penetration areas to withstand the inffects of an actuation of
, fire protection sprinklers or, in the. case of the. inner penetration,
'a loss of coolant. accident (LOCA). Splices selected for taping were
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those found in cables necessary to operate',the reactor during a LOC . The tape chosen was selected from seveial sutgitted to a testing lab for environmental testingtf Taping activity was inspected in ,
Section 10 of Inspection Report No. 50-155n8-1 ~
- ! 'f No violations or deviations were identified in}this area'. Monthly Maintenance Obsdrvation
Station maintenance activities of safety-related systems and components listed below were observed / reviewed to ascertain that they were conducted ,
in accordance with approved procedures, regulatory guides and industry codes or standards and in conformance with technical specification The following items were considered during this review: the limiting conditions for operation were met while components or systems were removed from service; approvals were obtained prior to initiating the
! work; activities were accomplished using approved. procedures ahd were
! inspected as appitcable; functional testing and/or calibrations were, i performed prior to returning components or'systetts to service; quality l control records were maintained; activities were accomplished by 7 l qualified personnel; parts and materials used were properly certified; radiological controls were implemented; and, fire prevention _ controls were implemente ., ,
Work requests were reviewed' to determine status of outskinding jobs and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc , [ During the week'of December 1, the inspector observed electrical wiring work performed in preparation for installation of the cleanup domineralizer bypass, During the week of December 1, the inspector observed portions of
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licensee activities to fabricate "
a foundation for the new diesel '
fire pump (DFP) driver ,
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c.- On January 3, the inspector d; served portion of licensee activities
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to remove the re:tetor plug,' core spray nozzle to the reactor head,
- and thermal shield. All activities were procedurally controlled and
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t involvement of radiological protection technicians was evident'.
While touring in the plant, the inspector noticed two postings of high radiation areas which were inappropriate and requested licensee corrective action. One incorrectly posted area was the turbine,-
which was shutdown, and the other was the general area where the reactor plug would be stored. .That area had a high radiation area posting sign lying next to floor tape bounding the area. That sign was removed and the turbine posting was covered ove , On January 3, the inspector observed portions' of licensee activities
to establist ah accurate data base for all valves in the plant in conjunction with a comprehensive valve packing control progra .The program works closely with a . single packing supplier to minimize future pa:: king problems by repacking valves with packing appropriately sized and qualified for the application. The program does not involve an immediate change out of existing valve packing but will replace packing as required. The licensee's efforts to' establish a data <
base involved physical measurements of valve stem and packing glands' ,
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plus do'cumentation of valve name plate dat ' On banuary 4, the inspector observed detensioning of reactor vessel head studs and removal of the reactor vessel head. The evolution was procedurally controlled, although in one instance a completed activity had not been signed off on the prccedure before proceedin Involvement of radiological control technicians and concern for ALARA prfaciples were evident. Examples included periodic work stoppage to
. read dosimeters, use of continuous air monitors and air samplings
" when the vessel head was lifted, and monitoring of radiation levels
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near the vessel head when stored. Prior to the evolution, a special training session was conducted by an instructor, who in his previous assignment,as a maintenance mechanic, participated in head removal several time J OnUJa5uary 4, the inspector observed portions of removal of Reactor Feed Pump No. I motor for shipment to an offsite repair
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facility. The inspector verified that tagged isolation for the
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' ' pump was adequat Radiological controls were in effect for the
, low level contamination present on the-inside of the motor housin On January 8, the inspector observed decontamination activities to repair the motor for shipment offsite. The licensee used a large steel pan normally used in fire training and lined it with plastic- s to establish an area where hydro-lancing of the motor could be performe On January 5, the inspector reviewed licensee plans to replace the diesel fire pump (DFP) driver. -The facility change, which will affect only the engine and not the pump end, was scheduled to commence after all fuel was removed from the reactor when core spray
_ water normally provided by the fire water system would no longer be
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irequired. During the scheduled two. week changeout, the electri fire. pump (EFP) would provide; fire suppression water for the
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i facility. The plant's backup firewater source during that period would be the Charlevoix Fire Department. Capability of the EFP to deliver the 1000 gpm required by Technical Specification 3.7.11.1-was verified by performance of Surveillance TR-70, Fire Suppression Water System Functional Test and Pump Capacity Test, last performed on October 8,1985, within the 18 month requirement of Technical Specifications 4.7.11.1. The procedure prepared to control the engine replacement includes requirements that no fuel movement will occur until both fire pumps are returned to service and declared operable. The overhaul schedule specified Surveillance TR-70 be performed in advance of fuel reloads. The licensee submitted to Region III the special report required by Technical Specification 6.9.3 and 3.7.11.1 when a firepump is to be out of , service beyond seven day '
Site Quality Assurance performed a special in plant surveillance to monitor the modificatio Throughout the remainder of the inspection period, the inspector observed the removal of the old engine, foundation and coupling
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modifications, and installation of the new driver. The activity
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was procedurally controlled and vendor contractors performing the work seemed generally knowledgeable of the need to adhere to procedural requirements. On one occasion, the inspector noted that'several procedural steps which had been performed correctly and in sequence were not signed off -on the working copy of the procedur Because of the location of the equipment, no radiological or contamination control measures were required. The inspector observed proper fire protection measures, including fire watches equipped with extinguishers during cutting operation On January 8, the inspector observed portions of overhaul activity on the target rock reactor depressurization valves in the Reactor Depressurization System (RDS). The work addressed lapping and polishing activities for main seat and disc on each of the four valves. The licensee suspected leakage past one or more of the main seats to be a major contributor to the relatively high unidentified leak rate observed during the closing months of the recently completed cycle. The work activity showed appropriate radiological controls were in place. The activity was procedurally centrolled but the inspector raised concern with licensee management over craft practices related to choice of grit in lapping compound and use of blue dye for performing disc to seat fit checks. Although the procedure calls for the application of lapping compounds of specific grit numbers, the term "or equivalent" is used. The t
inspector learned that polishing compound with a grit of 300,000,
^an extremely fine composition, had been substituted for the 8,000-grit compound specified in the procedure. The required grit was
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available on site and was provided when the project engineer learned of the substitutio l 1'
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The inspector learned that blue dye testing was not required by.the '
procedure and, although normally performed by repairmen as a craft practice, was not being performed on the main seats and discs. The
, inspector discussed with licensee management the benefits associated with blue dye testing early in-the lapping procedure to minimize.
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! unnecessary lapping vith heavy grit compound. The absence of dye:
- testing of seat to disc fit at the conclusion of lapping indicates that the work is completed and_ valve closed up- and-returned to-operability with no verification of true fit. The licensee performed .
! QC observed blue dye tests on all four valves. Site QA performed an in plant surveillance to monitor the modificatio . On January 9, the inspector reviewed Facility Changes (FC) 591,-
a plant modification to install. redundant-drain and vent valves
[i on the scram dump tank. The original, design uses only one vent i and one drain valve controlled through the Reactor Protection System, and the modification-will install a second valve in parallei
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concern over lack of redundancy registered in Section 5.3.8.2 of. the Integrated Plant Safety Assessment Report (NUREG-0828), which notes 3 the risk of a Loss of Coolant Accident (LOCA) via the scram dump tank-assuming a single failure of vent / drain valves. A probabilistic risk assessment (PRA) done in conjunction with a cost / benefit analysis of the proposed modification concluded that the redundant configuration !
would reduce the likelihood of a LOCA via failed vent / drain valves ,
by a factor of ten. The valves, which are activated by_RPS circuitry ,
on each plant scram, have no history of failure. The new valves will operate in unison with existing valves and be controlled by the same
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circuitry.
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The inspector's review of the FC concluded the modification was well <
decumented and researched. Procurement specifications, Q-list'
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, determinations, and receipt inspection and controls were in order :
and included appropriate certifications of conformance to code standards from the vendors. The piping and' structural analysis was
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i detailed, and a human-factors study was concucted to review necessary
alterations to the indication on the main control panel. -Position indication was provided for each of the four valves.
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A' review was '
conducted to modify procedures as appropriate, including the testing requirements of Technical Specifications to verify closure tim >
Drawing changes were provided, and Plant Review Committee (PRC) {
and site Quality Assurance each approved the modification.with minor comments. Maintenance orders to perform the modification were in draft form at the time of the inspector's review. Site QA has scheduled a special surveillance to monitor the modification In conjunction with the inspector's review, the licensee's proposed Technical Specification Change Request submitted to NRR December 30, 1986, was reviewed. The modification required inclusion of the valves
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in the operability tests of Section-5.2.2 h. .The inspector verified with the licensee the commitment made in the submittal to. implement I
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the proposed change following completion of the modification even though the change may not have completed all required steps in the NRR approval process for Technical Specification Changes by the time the modification is complet During the week of January 26, the inspector observed modifications to the control room console to install new valve position indication lights required by the modificatio On January 10, the inspector observed the installation of 480V three
~ phase breaker protection in an existing switchgear. The breaker, which will protect the new spent fuel pool filtration system, was installed with the switchgear energized. Electrical shock safety precautions for workmen were in evidence, including rubber deck mats, glove and arm protectors, and rubber insulation between phases within the switchgear. Supervisory participation was observed. Barriers for rerouting others away from the switchgear's high traffic location were not in place until .the final stages of the wor The Plant Safety Committee has recently placed higher emphasis on . electrical safety among maintenance pe'rsonne k. On January 12, the inspector observed tagging and isolation activity necessary to prepare shutdown heat exchanger "B" for gasket repair on the heat exchanger's shell side. Plant operators originally hung tags for removal from service as specified by System Operating Procedure (S0P) No. 5 which includes closure of the heat exchange vent Valve VSC-104. An alert auxiliary operator later observed that the tag hung on VSC-104 would not provide the necessary isolation for the intended work'and alerted control room operators. The original tag order was cancelled and a new isolation scheme-established and tagged, likely avoiding a spill of contaminated water when the heat exchanger was disassembled for maintenanc At the close of the period, the licensee was considering revision of S0P . Beginning January 13, the inspector observed the progress of Facility Change (FC) 599B, Source Range Neutron Monitoring. The log rate count meters associated with the old system were removed from the control room backpanel and the replacement Source Range meters and pre-amplifiers on startup Channels 6 and 7 were installed-in their place. Cable pulling activities in.the control room and sphere required breaking the integrity of the fire barrier in the control room floor and outside cable penetration area, and the licensee commenced hourly fire patrols in those areas. All work was controlled by procedure. The inspector discussed with the licensee the lack of QC inspector involvement and verify correct electrical connections prior to operational testing. The licensee's safety evaluation concluded that the source range was not a safety related item because (1) it provided indication and short period annunciation only and sent no signals to the Reactor Protection System (RPS), and (2) was not essential to mitigate the effects of an accident. The inspector
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- expressed the concern that although no RPS trip functions were-l generated by source range it.was essential to safe operation-of
the reactor during startup, since inoperability or inaccuracy of the instrumentation would result..in startup ofLthe reactor in a ,
possibly unmonitored condition.until neutron' count level reache i the. intermediate range. The inspector concluded that. site QA has adequately addressed this concern by scheduling a .QA walkdown _ i surveillance on the entire. installation prior to having~the system
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declared operable by the Shift-Supervisor. The operational test'
program on the FC prior to declaration of operabili.ty will involve-readout of actual. neutron level from a-neutron source and will be conducted prior to return of fuel-to the vesse On January 17,-the inspector observed cable pulling activities within
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the sphere for installation of new source range instrumentation on Channel '
On January 26, the inspector reviewed the details of the QA
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., surveillance and found it.to be thorough and detailed. . On January 26-27, the inspector observed the commencement of operational acceptance testing utilizing vendor resource Results of the test'-
program will be presented in a future inspection. repor ' On January 15, the inspector observed portions'.of the disassembly for grease change out of Limitorque Valve M0-7058, Reactor Water ;
from Shutdown System isolation valve. The. work was procedurall ~
controlled with appropriate measures to limit'possible spread of
- contamination. The valve was removed from the area where it is,
- normally installed to minimize worker exposure.
l On January 16, the inspector observed portions of the control: rod drive filter changeout using Procedure MCRD-6.
l On January 15-17, the inspector observed the disassembly.of Limitorqu Valve M0-7059, Reactor Water Isolation from the Shutdown System, for
- grease changeout and condition inspection. Appropriate measures-for
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contamination control were in evidence and the work was performed using procedural guidanc During the period January 16-20, the inspector. observed the handling i and shipping of three of seven planned shipments of radioactive' waste i for burial. The shipments consisted of old fuel channels previously.
- crushed and prepared for shipment. The evolutions were procedurally
! controlled, although the procedure was modified several times during j the course of the activit Participation of the. Project Engineer was extensive, and radiological and contamination control measures were in evidence throughout.
- During the transfer to the reactor deck of 'a.small cask on January 16,.
i the inspector observed activities and expressed the following concerns
- to the licensee
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(1) The view of the crane operator moving the small ' cask was l obstructed by a large cask positioned on the; reactor. dec Unable to see the individual' directing his activities via
- hand signals, the crane operator failed to stop crane travel 5 immediately -upon . the first . signal . .When the controller moved ;
L to be within view of the crane operator a'nd the signal was-. >
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observed, crane travel halted.
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(2) System Operating Procedure (SOP) 44, Spent Fuel Pool Operations t
and new Fuel Handling, prohibits the movement of. casks other_~ .
- than the fuel transfer cask over cr near stored spent fue Additionally, the SOP restricts all-cask operations other than those involving the fuel transfer cask, to the southwest corner
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of the spent fuel pool. The inspector observedithat after the small cask was positioned adjacent to the southwest corner.of
- the fuel pool on the reactor deck operators transferred the a cask to the northwest corner and placed _it on the reactor deck .
adjacent to the fuel. pool. .That movement required cask travel
! over the fuel pool and racks containing fuel removed from the reactor. The inspector's concerns were promptly addressed.
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, (3) When the small cask was initially placed adjacent to the fuel 4 pool's northwest corner, operators. identified the need to. move
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the reactor crane eastward in order to deenergize.the control
, rod drive interlocks associated with reactor crane. position and
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thus facilitate control rod drive maintenance activity,in another
- area. When operators began to slacken the lifting slings;on the t
cask and move the crane eastward, the-inspector pointed out the ,
- possibility of tipping the cask over into the fuel pool and onto !
the fuel stored in racks below. Operators abandoned the approach i j and removed the cask from the reactor crane before moving the-i crane to release the interlock.
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(4) The inspector reviewed the question of crane operator qualification with the licensee after observing the crane
- operator was a member of traveling maintenance crew and not normally assigned te Big Rock Point. ~The inspector expressed
- a concern that the crane operator may be a competent crane l operator, but unaware of the_special load control and crane
- travel requirements associated with work in the vicinity of
- the reactor and spent fuel pool. The licensee committed to
. instruct maintenance supervisors to ensure that prior to crane
, operation by an individual not normally. assigned to Big Rock
. the special requirements for crane travel associated with the j planned work would be clearly conveyed to the' crane operator.
j At the close of the inspection period a fourth loaded shipping
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container was undergoing evaluation for cracked welds.
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q. During the period, the inspector performed inspection activities that monitored licensee response to selected concerns raised during a special safety inspection conducted September 15-19, 1986, that examined the licensee's program to establish and maintain electrical equipment qualification (EEQ) within the scope of 10 CFR 50.4 (Reference Inspection Report No. 50-155/86013(DRS))
Open Item 50-155/86013-11(DRS) discussed lack of licensee response to the concerns detailed in IE Notice 83-72~for Limitorque Operator Those concerns focused on terminal blocks, wiring, space heaters, and structural components on Limitorque operators which were known to be unqualified or of undetermined qualification. The inspector reviewed the motor operated valve inspection sheet used to perform inspections and found it gathered data sufficient to evaluate the individual operator against the concerns in the notice. Additional Limitorque inspection guidance was provided by the inspecto The inspector observed the following licensee actions to address the Open Item:
On January 10, the inspector observed EEQ inspection of Valve M0-7066, a Limitorque isolation valve for fire water to the core spray heat exchange In addition to verification that terminal boards, limit-switches, torque switches, switch rotors, and structural components were of the correct material, a general inspection was conducted to evaluate the operator's general condition. One insulation crack was located in internal wiring and relugged, and several internal wires were routed to avoid contact with other internal components. A grease sample was taken which revealed some discoloration of the approved lubricant with no grit observed. Torque switch settings were verified. The capacity and positioning of the internal heater was verified appropriate in response to concerns raised in'IE Notice 86-71 which alerted licensee to several instances where internal heaters caused insulation burning on adjacent wires. The licensee began a technical review to determine the effect of heater removal. The 25 watt heater is installed for use when the operator is stored and the licensee speculates it is not necessary for operators installed in a temperature controlled environmen Unresolved Item 50-155/86013-05(DRS) maintained that the operator for M0-7068, backup enclosure spray isolation' valve, was not qualified due to presence of unqualified materials and the licensee committed to operator replacement. On January 10, the inspector observed the removal of the operator and partial installation of a qualified replacement. Electrical hookup was delayed awaiting fittings and parts. The replacement was controlled with thoroughly prepared procedures. Prior to the installation the assigned repairmen walked through the job with their supervisor. Radiological controls were in evidence.
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On January 22, the inspector ob' served replacement of terminal lugs
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on wires within'the controller on Limitorque Valve M0-7068. :The inspector also observed-removal of _ old Ray Chem splice material
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and preparation of the motor leads for ~new connectors and splice ; On January 10, the inspector observed grease sampling activity on MO-7080, bypass valve for fire water-to core spray heat exchanger, an EEQ valve. Records: indicated the original grease was'two year old,> slightly thickened and of sticky consistency. No grit was l observe j On January 22, the inspector ob' served testing of' Control Rod Drive-Relief Valve RV-5051 under Surveillance Procedure T-365-19 (forme'rly:
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TR-75). The relief valve ~ lifted at'1803 psig, below the specified
- lift point 1950, plus1or minus 25 psig. The licensee decided to-disassemble and inspect the valve and-identified defects in seat
and disc. The repair was incomplete at the close of the inspection-period.
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l Throughout the. refueling overhaul, the inspector followed-licensee-4- activity to test steam drum relief valves. Testing is conducted-
- in accordance with the requirements of ANSI /ASME OM-1, 1981,.
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" Requirements for Inservice Performance Testing'of Nuclear Power i Plant Pressure Relief Devices" which defines test frequency and- 1 i required additional' testing when any valve in a system fails to !
function properly during a regular _ test. The first installed valve to be tested, No. A-6 with a lift point of 1585, plus or minus five
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(5) psig, actually lifted at 1822 psig. The valve could-not be repressurized for subsequent lifts and was considered to have faile'd, j thus necessitating testing of a second relief valve.. Relief valve j No. A-3 with a setpoint of 1535,.plus or minus five (5) lifted in- ,
i the as-found condition at 1707 psig. Subsequent-lifts were at 1496, '
i 1503 and 1501 psig and Valve A-3 was considered a failur In keeping with the code requirements all six reliefs were removed: for testin '
j Licensee personnel investigating the binding 'effect of the disc j and seat observed on the first two valves tested noted a-gray discoloration on the disc seating face. The investigation' focused on possible causes of chemical adhesion between~ disc and seat,.
possibly from lapping, polishing or cleaning chemicals. used during valve rebuild. A third valve, A-1 was carefully disassembled.to i leave the disc intact on the seat and pressurized to 153 psig before
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the disc separated from the seat. The disc was sent to an offsite
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laboratory for analysis, where it- remained at the close of,the
, inspection period. Rebuild and setpoint checks on all steam drum
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relief was placed on hold pending the laboratory results and ~
- corrective actions. The high lift pressures 'were _ reported to
- NRC via ENS on January 2 !
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During the review of steam drum relief valve testing, the inspectorL expressed a concern that PRC approved changes to-relief valve testing
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methods.and acceptance criteria in Surveillance TR-28, Steam Drum -
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, Relief Valve Set Point and Acceptance Determination, would effectively-
permit a relief valve that lifted well outside specification to be declared acceptable if it performed within. specifications.on two i subsequent lift attempts. The procedure change stated that "fo acceptable valve performance, data must demonstrate' repeatability and the last two measured valves must lie on/in the acceptance band."
The inspector's position emphasized theLimportance of the initial lift valve in determining the valve's acceptable performance, noting -
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that the initial "as-found" lift point. is very likely to be the pressure at which the re_ lief would have lifted had it been called- '
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upon to perform its relief function while still installed in the '
- system. A similar procedure change was made~in other tests for
- relief valves installed throughout the plant. The license based .
- their decision to change to the les's_ restrictive acceptance criteria
] on the contents of IEN 86-56, Reliability of Main Steam-Safety. Valves l and on their interpretation of OM-I, Article 4.1. ~
The question j was resolved by contacting several members of the ANSI /ASME OM-I i Code Committee who agreed that the committee's intent was that
"as-found" setpoint was the sole criteria to determine acceptable '
performance. Subsequent -lifts referenced in Article 4.1.1.9 are-used to establish repeatability after maintenance. On January 26,-
the licensee initiated procedure changes-that would make the 1
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"as-found" setpoint the sole determinator for acceptability.
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l No violations or deviations were identified in this this area.
, Surveillance Observation
- On December 17, the inspector-observed licensee quarterly surveillance
activities to successfully verify the accuracy of wind speed and
!. direction as indicated by station meteorological instrumentatio The test requires the application of constant speed motors to the
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wind speed instrument and a protractor device to the wind direction instruments, bcth of which are located atop the plant stac ;
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Recalibration of speed and direction readouts are then performed -
at the meteorological station in the Technical Support Cente o f' On December 23,-the inspector observed-portions of 'l Surveillance T30-31, Reactor Depressurization Cabinet Test,
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i on sensor cabinet D/ Actuation Cabinet 4. .The surveillance was-
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successfully performed in accordance with a detailed procedure.
- The inspector noted the conservative approach of operators when
- off-normal indications were received for reactor and steam drum levels. Operators immediately halted the surveillance and involved Instrument and Control Technicians to~ restore system indication l
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~ During the refueling period, the inspect'or observed all or part of the following surveillances. In.each case the surveillance was
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conducted.according to approved procedure based on an, inspector' review of the test documen (1) On January 5, the inspector observed the' performance ~of- -
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Surveillance TR-64, Nut and Bolt Surveillance Test. .The test involved the use of a remotely controlled miniatur closed circuit television camera to-take-inventory of all mechanical hardware and fasteners located above fuel.
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(2) On January 6, the inspector observed portions-of .
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Surveillance TR-96, Control ~ Rod Withdrawal Interlock
- Tes (3) On January 8, the inspector observed portions of the-t performance of TR-23, Neutron Source Removal, Inspection, and Reinsta11ation.
v l (4) On January 9, the inspector observed portions'of th performance of.TR-69, Fire Penetration Barriers, Nozzle
and Hose Inspection. The' integrity of the fire barrie formed by the south wall of the condensate pump room was compromised by a circular hole through'which passed-a piping run, resulting in the implementation of hourly fire-patrols
in that area. Several discrepancies were discovered durin the surveillance in which lights, pipes, or structural components were closer than the-eighteen inch: limit to a sprinkler nozzle, creating-the possibility of obstructing i sprinkler flow. Corrective. action was incomplete at the close -
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(5) On January 12, the inspector observed portions of TR-65B, !
Reactor Depressurization System Uninterruptible Power Supply- .,
i- "B"' Battery Service Test and Discharging Alarm Operability.
! Verificatio (6) On January 16, the inspector observed portions of TR-39F, Containment Isolation Valve- Leak Rate Test for Treated Waste
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(7) On January 16, the inspector observed. portions o Surveillance T-30-40, Reactor Crane Survey. The inspector ,
i noted the satisfactory performance of the daily crane checklist '
associated with the procedure.
! (8) On January 21-23, the inspector observed portions lof-Surveillance TR-84, Emergency Diesel Generator Inspection,
.' conducted by plant. maintenance staff and a vendor l representative. The Surveillance involved inspection t
electrical and mechanical components, repair of oil-leaks, and antifreeze changeout.
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(9) On January 22, the inspector reviewed the results of. chemical
, analysis for sodium pentaborate in the plant's liquid poison
! system as required by Surveillance TR-04. During the previous cycle, licensee auditors had generated an action item request-to address the concerns of IE Information Notice 86-48, Inadequate. Testing of Boron Solution Concentration In the Standby Liquid Control System. The licensee analyzed five samples,'each of which was subjected to an increasing amount of agitation with air to minimize stratifcation or separatio In each case, the samples were between the limits of 19 to 30 weight percent sodium pentaborate required by Technical Specification 5.2.3, with 28.2 weight percent as a representative readin (10) On January 22, the inspector observed portions of TR-12 and TR-11, Liquid Poison System Explosive Valve Bench Tes Technicians used monitored input firing current to activate the explosive charge in a test stand to test seven squib valves-removed from the system after five years of servic In addition, one squib valve was tested from among new
- replacement valves recently received from the manufacturer as a test of the quality of
- quib valves being stored or installed during the next-five year period. The test method met the requirements of Technical Specification 5. for shearing of the integral inlet cap (11) On January 23, the inspector observed portions of Surveillance IFPS-7, Calibration of Diesel Fire Pump Instrumentation. Procedural control and appropriate tagged isolation were in evidenc . Training
' During a meeting between the Chem / Health Physics Superintendent and the inspector to discuss activities conducted over the previous months, the licensee listed as an objective the need for refresher training for advanced radiation workers. Noting that operators who
- receive training once a year may forget some requirements, the Training Department has integrated refresher training into the five week requalification training cycle. The Superintendent also noted that increased involvement of management and first level supervision has resulted in better communication with technicians and provided access to management for clarification of issue On January 21, the inspector observed training of control room
! operators on the new instrumentation installed under the facility change to replace source range nuclear instrumentation. Training used a lesson plan with specific objective and was conducted on an actual spare instrumentation modul ._ _ _ _ .__
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. On December 15, the inspector reviewed with the licensee the publication of a training manual for Field Maintenance Service (FMS) and Contractor Supervisor orientation. The manual was designed as a reference document and supporting text for two days of plant orientation training. The manual was comprehensive and included a review of problems which led to escalated enforcement action during the 1985 outag On December 16 and 17, the inspector observed portions of plant orientation training conducted for FMS and contractor supervisors in preparation for the 1987 refueling outage. The training sessions in the following subjects were taught by cognizant plant personnel who will serve as primary contacts for FMS and contractor personnel:
Maintenance Administration, Material Control, QA/QC, Health Physics /ALARA, Security / Fire Protection, Safety, and the role of-the newly created Planning and Scheduling Department. The two day session concluded with a plant tour conducted by the plant's primary
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contact individual. The training emphasized recent revisions to use of Radiation Work Permits, stressed the importance of strict adherence to plant procedures, and clearly conveyed the licensee's intention to avoid problems experienced in previous outages through job preplanning and close coordination of FMS activities through a plant primary contac No violations or deviations were identified in this are . Licensee Action on IE Bulletins By letter dated January 16, the licensee was informed by NRR that the staff's evaluation of additional data regarding calculation for temperature rise in the recirculation pump room due to an oil fire indicated that the conclusions reached in the original evaluation remain valid. An exemption from 10 CFR 50, Appendix R, Section III.0 for an oil collection system remains ir, effec No violations or deviations were identified in this are . Licensee Event Reports Followup Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with technical specification By letter dated January 23, the licensee submitted LER 87-001, Reactor Trip, Upscale /Downscale, an event reportable under 10 CFR 50.73. The LER describes the reactor trip detailed in Sections 2 and 8 of this ,
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O 8. Reactor Trips On January 2, during a normal shutdown to enter a refueling outage the reactor scrammed when erratic picoammeter behavior in power range instrumentation caused an upscale /downscale reactor protection system trip. The scram is discussed in Section 3 of this report. The reactor-at the time of the trip was at 480 F, 630 psig, and less than .001% reactor power, indicating the shutdown condition of the plan The inspector reviewed the findings of the Corrective Action Review Board on January 5 and January 15. The Board concluded that short term corrective action should include repair and/or replacement of picoammeter components as required. Long term corrective action focused on modifications to replace existing power range and intermediate range instrumentation with state of the art wide range instrumentation during the next refueling outag . Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)
throughout the month and at the conclusion of the inspection period and summarized the scope and findings of the inspection activities. The licensee acknowledged these findings. The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspectio The licensee did not identify any such documents or processes as proprietar l
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